Text

Amends 78,64 & 50 to Licenses NPF-41,NPF-51 & NPF-74, Respectively.Amends Change TS in Response to Application
	ML17310B462
Person / Time
Site:	Palo Verde
Issue date:	07/15/1994
From:	Quay T; Office of Nuclear Reactor Regulation
To:	;
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ML17310B463	List: ML17310B462; ML17310B464;
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON.'D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.

DOCKET NO. 'STN 50-528 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment 'No. 78 License No.

NPF-41 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 August 5,

1993, as supplemented by letter dated January 19,

1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended

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

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted wi,thout endangering the health and safety of the public, and (ii) that such activities will'e 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-41 is hereby amended to read as follows:

9aow& o~ss 9oo7i s PDR ADOCK 05000528 PDR

P r

I I't

~0

(2)

Technical S ecifications and Environmental"'Protection Pl'an 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 facility in accordance with the Technical Specificati'ons and the Environmental'rotection

Plan, except where

,otherwise stated in specific license conditions.

3.

This license amendment is effective as of,the date of issuance and must be fully implemented.

no later than 90 days.'from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance.

~uly i5:,'994 Theodore R.

Quay, Di'rector Project -Directorate IV-2 Division of 'Reactor Projects II-I/IV.

Office of Nuclear Reactor Regulation

P~

J

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

78 TO FACILITY OPERATING LICENSE NO.

NPF-41 DOCKET NO.

STN 50-528 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

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

Remove Insert 2-3 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-25 3/4 3-31 3/4 3'-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1 2-3 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-25 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1

0 0

if 1

FUNCTIONAL UNIT 1.

TRIP GENERATION A.

Process TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TRIP SETPOINT ALLOWABLE VALUES 1.

Pressuri zer Pressure High 2.

Pressurizer Pressure Low 3.

Steam Generator Level Low 4.

Steam Generator Level High 5.

Steam Generator Pressure Low 6.

Containment Pressure High 7.

Reactor Coolant Flow Low a.

Rate b.

Floor c.

Band 8.

Local Power Density High 9.

DNBR Low-B.

Excore Neutron Flux l.

Variable Overpower Trip a.

Rate b.

Cei ling c.

Band

<<2383 psia 1837 psia (2) 44.2; (4)

<< 91.0'.

{9)

~ 919 psia (3)

<<3.0 psig 0.115 psi/sec (6) (7)

~ 11.9 psid (6) (7)

<<10;0 psid (6)(7)

<<21. 0 kW/ft (5) 1.30 (S) 10.6'./min of RATED THERMAL POWER (8)

<<110.0>> of RATED THERMAL POWER (8)

<<9.7% of RATED THERMAL POWER (8)

<< 2388 psia

~ 1821 psia (2) 43.7>>o (4)

<< 91.S-.

(9)

= 911 psia (3)

<< 3.2 psig 0.118 psi/sec (6)(7) 11.7 psid(6) (7) 10.2 psid (6)(7) 21.0 kW/ft (5) 1.30 (5) all.0%/min of RATED THERMAL POWER (8)

<<111.0% of RATED THERMAL POWER (8)

<<9.9>>o of RATED THERMAL POWER (8)

FUNCTIONAL.UNIT'ABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TRIP SETPOINT ALLOWABLE VALUES 2.

Logarithmic Power Level - High (1) a.

Startup and Operating b

Shutdown C.

Core Protection Calculator System 1

rPh I s1nII1sfnve 4 ~

VI < I VIA ~ \\ %l

~ %l I V ~

2.

Core Protection Calculators D.

Supplementary Protection Sjsiem Pressurizer Pressure High Vr One Ihr VI Il ~

Br 4 LVQJ'4 A.

Matrix Logic B ~

InltI ation Logic III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manual Trip A 0101, o'f PATED THERMAL POWER U.ULU'h Ot KAltU Ma4 Anal'l IIV4 tllIlII II IIVI0 Not Applicable

< 2409 psi a, Not Applicable Noi Applicable Not Applicable Not Applicable h hl 1V C nIITefl V ~ Vtl/0 V I IV% I LIJ THERMAL POWER

< 0I011X of, RATED ToehUAI hnl.ll h IIILhIWL rVIICZ Not Appl 1caole Not Applicable

< 2414 psia Not ADDlicable Not Applicable Not Applicable Not Applicable

THIS PAGE INTENTIALLYLEFT BLANK PALO VERDE - UNIT '1 3/4 3-13 AMENDMENT NO. 24

TABLE 4.3-1 REACTOR PROTECT V

NSTR N

I N SURVEILLANC RE U

REM NTS FUNCTIONAL UNIT I.

TRIP GENERATION h

Dv nr ace R

R ~

~

~ V%

1.

Pressurizer Pressure - High DRRRAI 8 ~ RN4%hRRR ANAiR ~ ~

I C

~

I I VDBlll I C.VI I I 'C>DMI 'C LVw 3.

Steam Generator Level Low 4.

Steam Generator Level - High CHANNEL CHANNEL CHECK CALIBRATION CHANNEL MODES IN IIHICH FUNCTIONAL SURVEILLANCE EE

~EEIRE 1,

2 Ig 2

1, 2

5.

Steam Gener ator Pressure Low 6.

Containment Pressure - High 7.

Reactor Coolant Flow - Low 8.

Local Power Density - High 9.

DNBR - Low D

fur ave kl

~ ~ 4 C1

~ ~

0 ~

CA'VVI '4 lltlllrlVII I IUA 1.

Variable Overpower Trip D (2, 4),

R (4, 5)

D (2., 4),

R (4,R 5)

M (8),

S (7)

D (2, 4),

M (3, 4) 0 -(4) 0 0

0, R (6) 0, R (6)

1. 2. 3*. 4*

1 2

1, 2

2.

Logarithmic Power Level - High S

C.

Core Protection Cal,culaior System 1.

CEA Calculators 2.

Core Protection Calculators R (4)

D (2, 4),

R (4, 5) n goy, g and S/U (1)

I, 2, 3, 4, 5

and

~

R (6)

0. (9),

R (6) 1, 2

TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIRENENTS U

I CD FUNCTIONAL UNIT 0.

Supplementary Protection System Pressurizer Pressure

- High II.

RPS LOGIC CHANNEL CHANNEL CHECK CALIBRATION CHANNEL NODES IN WHICH FUNCTIONAL SURVEILLANCE

. IEE.

~UIII 1,

2 A.

matrix Logic B.

Initiation Logic III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Hanual Trip N.A.

N.A.

M, R (10) 1, 2, 3*, 4*, 5*

3*

4*

5*

1, 2, 3*, 4*, 5*

3*

4*

5*

CD m

CD

(3)

(4)

(5)

(6)

(7)

(8)

(g)

(10)

TnBCE 4.3-1 (Continue~d ~

IIABLE NO I ATIOtLS With reactor trip breakers in the'losed position and the CIEA drive system capable of CEA withdrawhl, arid fuel in the reactor vi ssel Each STAR1I'UP or when required wit'h t'.he'e'actor trip breakers closed and the CEA drive system capab'le of rod withdrawal, if not Iperformed in the. previous, 7

days'

Heat balance only (CHANNEIL FUNCTIONAL t'EST riot'included),

aboVe

!l5X'f RATED THEIRMAL POWER; adjust the linear power level, ttie CPC delta 1'ower and CPC nuclear jiower si(jnals to agree with the calorimetric calculatipn if absolute difference is I~reater than 2X.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon rbacthihg each major test power plateau and prior to.proceeding to the nekt major test power. plateau.

Above 15X of RATED THERMAL POWER, verify thait the.linear power sub-channel gains of the excore detector's are consistent with tlute values used to establish the shape ankeaIlinlg thatrix elements in thIa Core Protection Calculators.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

After each fuel loading and prior to exceeding 70X of RATED THERMAL

POWER, the incore detectors shall be used, to determine the

. hape annealing matrix elements and ithe Core PrOtection Calculators shall use these elements.

This CHANN'EL FUNClIOINAL TEST shall incllude the injection of siimulatiad process signals into the channel as close to the sensors as pract.icable to verify OPERABILITY including alarm an'd/or trip functions Above 70X of RATED TfciERMAL POWER, verify that the total steady-state RCS flow rate as indicated by dach CPC is less than or equal tIo tthe actual RCS total f'low rate detkrmIined by Lit'her using the r(~actor coolant pump differential pressure instrumentation or by calor~imdtric calculations and if necessary, adIjust the'PC. addressable constant flow coefficients such that each CPC indicated flow. is less'th'an

'or'qual to the actual flow rate.

The flow measurement, unce'rt0inty 'mag be included in the BIERRl term in the CPC and is equal to or greater~

than 4X.

Above 70X of RATEiIi THERMAL POWER; verif'y that the total steady-state

RCS f'low rate as indicated by each CPC is less than or equal tp the actual RCS tota'1 f'low rate determined by either using the, reactor'oolant pump differential pres0ur0 instirumen'tation and the ultrasonic flow meter adjusted pump curves or calorimetric calculations.

The qIuarterly CHAN'NEI FUNCTIONAL TEST shall include verificdtibn that the correct current values of addressable'cohstants are installedin each OPERABLE CIPC.

At least once per 18 months and following maintehance or adjustment of the reactcir trip breakers, t'he'CHANNEL'FUNCTIONAL TEST shali include indetiendent verification 6f the undervoltage

.and shunt trips.

PALO VERDE - UNIT 1

3/4 3-16 AMENDMENT NO. -'tP;78

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES m

Kl Cl m

ESFA SYSTEM FUNCTIONAL UNIT I.

SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units 1.

Containment Pressure High 2.

Pressurizer Pressure Low B.

ESFA System Logic C.

Actuation Systems II.

CONTAINMENT ISOLATION (CIAS)

TRIP SETPOINT

< 3.0 psig

> 1837 psia Not Applicable Not Applicable ALLOWABLE VALUES

< 3.2 psig

, > 1821 psia"'ot Applicable Not Applicable C.

A.

B.

A.

B.

C.

Sensor/Trip Units 1.

Containment Pressure High 2.

Pressurizer Pressure Low ESFA System Logic Actuation Systems CONTAINMENT SPRAY (CSAS)

Sensor/Trip Units Containment Pressure High - High ESFA System Logic Actuation Systems

< 3.0 psig

> 1837 psia"'ot Applicable Not Applicable

< 8.5 psig Not Applicable Not Applicable

< 3.2 psig

> 1821 psia'-

Not Applicable Not Applicable

< 8.9 psig Not Applicable Not Applicable IV.

MAIN STEAM LINE ISOLATION (MSIS) m Cl m

CO A.

B.

C, Sensor/Trip Units 1.

Steam Generator Pressure - Low 2.

Steam Generator Level High 3;

Containment Pressure High ESFA System Logic Actuation Systems

> 919 psia'

< 91.0'R

< 3.0 psig Not Applicable Not Applicable

> 911 psia'

< 91.5X NR' 3.2 psig Not Applicable Not Applicable

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION.TRIP VALUES ESFA'SYSTEM FUNCTIONAL UNIT V.

RECIRCULATION (RAS)'.

Sensor/Trip Units Refuelinn Mates Storane Tank - Low B.

ESFA'ystem Lo'gic C.

Actuation System VI.

AUXILIARYFEEDMATER '(SG-1)(AFAS-1)

A.

Sensor/Trip Units 1.

Steam Generator 81 Level - Low 2,

Steam

-Generator h Pressure-SG2

> 561 B.

ESFA System Logic'.

Actuation Systems VII. 'AUXILIARYFEEOMATER (SG-2)(AFAS-2)

A.

Sensor/Tlip Units 1.

Steam Generator N2'evel - Low 2

Steam Generator h-Pressure-SG1 SG2 B.

ESFA-System Logic C.

Actuation Systems VIII.

LOSS OF POMER A,

4.16 kV Emergency Bus Undervoltage (Loss of Voltage)

B.

4.16 kV Emergency Bus Undervoltage (Degraded Voltage)

IX.

CONTROL ROOH ESSENTIAL FILTRATION TRIP VALUES 7.4X of Soan Not Anni 1 cab 1 e II j A

11 L1 not, nppiicauie

> 25.8X MR 4

< 185 psid Not Applicable Not Appl'icabie

> 25.8X MR(4)

< 185 psid Not Applicable Not Applicable

> 3250 volts 2930 to.3744 volts with a 35-second maximum time de iay

< 2 x 10-s pCi/cc ALLOWABLE VALUES 7.9 > X of Span

> 6.9 Not Applicable

'l I A~al i~~L1 a IIV'lgwuI Il OUI'6

> 25.3X MRi"

< 192 psid Not Appl1cabl o ll I A~~7 ir +ADra llo4 tip@ I Il Oll I K

> 25.3X MRi"i

< 192 psid Not Applicable Not Applicable

> 3250 volts 2930 to 3744 volts with a 35-second 2 x 10-s pCi/cc

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREHENTS ESFA SYSTEM FUNCTIONAL UNIT

- High

- Low ys)

- High

- Low SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units 1.

Containment Pressure 2.

Pressurizer Pressure B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logic 3.

Manual SIAS C.

Automatic Actuation Logic (except subgroup rela Actuation Subgroup Relays CONTAINMENT ISOLATION (C IAS)

A.

Sensor/Trip Units 1.

Containment Pressure 2.

Pressurizer Pressure B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logic 3.

Manual CIAS 4.

Manual SIAS CHANNEL CHECK NA NA NA NA NA NA NA NA NA CHANNEL CALIBRATION NA NA NA NA NA R

R NA NA NA NA CHANNEL FUNCTIONAL TEST 0(2)

H(1) (3)

MODES FOR WHICH SURVEILLANCE IS RE UIRED 1, 2, 3, 4

1, 2; 3, 4

1, 2, 3, 4

1; 2, 3, 4

1, 2, 3

1; 2, 3, 4

1, 2, 3, 4

TABLE-4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREHENTS m

X7 C7m C

ESFA SYSTEM.FUNCTIONAL UNIT II.

CONTAINMENT ISOLATION (Continued)

CHANNEL CHECK CHANNEL CALIBRATION CHANNEL MODES'OR MHICH FUNCTIONAL SURVEILLANCE E

~EII E

C.

Automatic Actuation Logic (except 'subgroup relays)

Actuation Subgroup Relays I II.

CONTAINMENT'PRAY (CSAS)

. NA NA NA NA 0(2)

H(1) (3) 1, 2,

3,,

4 A.

Sensor/Trip Units Containment Pressulre--

High High ESFA System Logic 1.

Matrix Logic 1, 2, 3

1, 2, 3, 4 C ~

Inltiation Logic 1,2,3,4 3.

Manual CSAS 1.,2,3,4'

~

AutomEEatic= Actuation Log)c (except subgroup relays)

Actuation Subgroup Relays I g EP I gg)

H(1) (3) 4 1, 2, 3, 4

vi)

~

m O

'V SGi > Sls a t4 A"'~~

$ S ~. ~ ~

c'zc'a l,il ILr n<<.l

TABLE 4.3-2 (Continued)

ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATION SURVEI LANCE RE UIREHENTS ESFA SYSTEH FUNCTIONAL UNIT IV.

HAIN STEAH LINE ISOLATION (HSIS)

A.

Sensor/Trip Units 1.

Steam Generator Pressure-Low 2.

Steam Generator Level

. High 3.

Containment Pressure High CHANNEL CHECK CHANNEL CALIBRATION CHANNEL HODES FOR MHICH FUNCTIONAL SURVEILLANCE ii i

~IE i Ui ii-1, 2, 3, 4

1, 2, 3, 4

1, 2; 3, 4

B; ESFA System Logic 1.

Hatrix Logic 2.

Initiation Logic 3.

Hanual HSIS NA NA 1, 2, 3, 4

1, 2, 3, 4

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays NA NA

'A NA Q(2)

H(l) (3) 1, 2, 3, 4

1, 2, 3, 4

TABLE 4.3-2 (Continued)

ENGIN ERED SAFETY FE TURES ACTUATION SYSTEH INSTRUHENTATION SURVEILLANCE RE UIREHENTS ESFA SYSTEH FUNCTIONAL UNIT V.

RECIRCULATION (RAS)

Cn~ena ITS

% II IIIIi+'C s>4IIPMI r I I I II Mll I I 4 Refueling Water Storage Tank - Low CHANNEL CHECK CHANNEL CALIBRATION CHANNEL HODES FOR'WHICH FUNCTIONAL SURVEILLANCE E

~IE EIREE 1, 2, 3

B, ESFA System Logic 1.

Hatrix Logic Initiation Logic 3.

Hanual RAS NA 1, 2, 3, 4

I 2, 3, 4

1, 2, 3, 4

VT Qgtomaticl Actuation I ogic (except subqroup relays)

Actuation Subgroup Relays AUXILIARY-FEEDWATER (SG=l)(AFAS=l)

A.

Sensor/Tri'p Units Steam GeneraR+tor 0'I Level Low 2.

Steam Generator

~ Pr cssur e SG2

> SG1 NA NA NA NA Q(2)

H(1) (3) 1, 2, 3, 4

1, 2, 3, 4

1,2;3 RI Lr Cg

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM.INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL ESFA SYSTEM FUNCTIONAL UNIT CHECK VI.

AUXILIARYFEEDWATER (SG-1) (AFAS-1) (Continued)

CHANNEL CALIBRATION CHANNEL MODES FOR WHICH FUNCTIONAL SURVEILLANCE TE T 15 EISUIREO B.

ESFA System Logic 1.

Matrix Logic 2.

Initiati on Logi c 3.

Manual AFAS NA 1, 2, 3, 4

1, 2, 3, 4

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays VII.

AUXILIARYFEEDWATER (SG-2)(AFAS-2)

NA NA NA NA O(2)

M(1 (3) 1, 2, 3, 4

1, 2, 3, 4

A.

B.

C.

VIII. LOSS A.

B.

Sensor/Trip Units 1.

Steam Generator P2 Level -Low 2.

Steam Generator h Pressure SG1

> SG2 ESFA System Logic 1.

Matrix Logic 2.

Initiati on Logi c 3.

Manual AFAS Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays OF POWER (LOV)

4. 16 kV Emergency Bus Under-voltage (Loss of Voltage) 4.16 kV Emergency Bus Under-voltage (Degraded Voltage)

NA NA NA NA NA

'NA NA NA S2)

M I) (3) 1, 2, 3

1,-2, 3

1, 2, 3, 4

I, 2, 3, 4

1, 2, 3, 4

1, 2; 3, 4

I, 2, 3, 4

1, 2, 3, 4

ll'ABLE NOTATION (1)

Each train or logic channel shall be tested at least every 62 days on a

STAGGERIED TEST BASIS.

(2)

Testing of'utomatic actuation. logic shall i,nclude energizatian/

deenergization,of each initiation rdlag andiverification of proper, operation of each initiation relay.

(3)

'A subgroup relay test sha'll be performed which shal'll include the energizatiion/deenergization of each subgroup relay and veri'fication of the OPERABILITY of eaclh subgroup relay.

Relays listecl below are exempt from, testing during POWER OPERATION but, shall be tested at, least once per 18 months I9uring REFUELING and-during each COLD SHUTDOWN condit,ion unless tested within the previou 62 days.

ACTUATION DEVICES THAT CANNOT BE TESTEG AT POWER TRAIN A ESF FUNCTION SIAS A SIAS A

CIAS A

'CIAS A CSAS A

HSIS A HSIS A

AFAS 1A AFAS 2A ACTUATION DEVICE K108 K409 K202 K204 K304 K305 K404 K211 K112 TRAIN 8 ESF I

FUNCTION SIAS 8 SIAS 8 CIAS 8 CSAS 8

HSIS 8

AFAS 18 AFAS 1'8 AFAS 28

'CTUATION

'EVI'CE'108 K409 K204 K:304 K:305 K404 K113 K211 K112 one or more pieces of equipnient cannot be actuated, but can be rackedi out, i

bypassed or etc.,

which wi111 not preclude the relay from being tested bUt will not actuate the locked out equipment associated witlh the relay:

SIAS A SIAS A SIAS A CIAS A CIAS A RAS A RAS A RAS A AFAS 1A K401 K410 K412 K203 K210 K104

.K312 K405 K113 SIAS 8 SIAS 8 CIAS 8 CIAS 8 RAS 8 RAS 8 RAS 8 K301 K308

'K203

'210 K104 K312 K405 PALO VERDE UNIT 1

3/4 3-36 AHENDMENT NO.

27

i~

4.3 INSTRUMENTATION BASES 3 4.3.1 and 3 4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETiY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of.the 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 will be initiated when the parameter monitored by each channel, or, combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

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

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

Ol.

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 factors and CEA deviation penalties.

Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8. 1) ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.

The 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.

PALO VERDE UNIT 1

B 3/4 3-1 AMENDMENT NO.~ 78

INSTRUMENTATION BASES~

'"':l

' UI INSTRUMENTATION (Continued)

The value of 'the DNBR in Specification

2. I i's conservatively compensated for measurement uncertainties,.

Therefore, the actual RCS total flow rate determined by the.-reactor coolant pump dif'ferential pressure instrumentation or, dy

'alorimetric calculations does not have to be conservatively compensat~ed~for-

'easurement uncertainties The measurement of'esponse time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed withlin the, t'ime limit assumed in the safety analyses.

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

The response times in Table 3.3-2 are made up of'he time to generate the trip signal at. the detector (sensor response time) and the time for the signal to i,nterrupt power to the CEA.drive mechanism (signal or trip delay time).

PALO VERDE UN:[T 1 B 3/4:I-2 AHENDIlENT NO.

24

- ~

~pe REC0 Wp

~

0 n 'P~

Ih Cy

~O UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ARIZONA PUBLIC SERVICE COMPANY ET AL.

DOCKET NO'.

STN 50-529 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No.

NPF-51 1.

The Nuclear Regulatory Commission (the Commi'ssion) has found that:-,

A.

The application for amendment by the Arizona Publ.ic 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 Authority dated August 5,

1993, as supplemented by letter dated January 19,.1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended'the Act) and the Commission's, regulations set forth in 10 CFR Part 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 compl'iance 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.

'ccordingly,,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 Faci.lity Operating License No., NPF-'51 is hereby amended to read as follows:

'P

~

(1

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained'n Appendix A, as revised through Amendment No.64 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 and. must be fully implemented no later than 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

Quly 15, 1994 Theodore R. quay, Director Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear. Reactor Regulation

li t.

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE.NO.. NPF-51 DOCKET NO.

STN 50-529 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

The revised -pages are identi-fied by amendment number and'ontain vertical lines indicating the areas of change.

Remove 2-3 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-25 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1 Insert 2-3 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-25 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1

gP 0

L

~

TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT. LIMITS FUNCTIONAL UNIT 1.

TRIP GENERATION A.

Process 1.

Pressurizer Pressure High 2.

Pressurizer Pressure Low 3.

Steam Generator Level Low 4.

Steam Generator Level High 5.

Steam Generator Pressure

Low'.

Containment Pressure High 7.

Reactor Coolant Flow Low a.

Rate b.

Floor c.

Band 8.

Local Power Density High 9.

DNBR Low B.

Excore Neutron Flux 1.

Variable Overpower Trip a.

Rate b.

Ceil ing c.

Band TRIP SETPOINT

<<2383 psia

~ 1837 psia (2)

> 44.2>> (4)

<< 91.0>> (9)

~ 919 psia (3)

<< 3.0 psig

<< 0.115 psi/sec (6)(7) 11.9 psid (6)(7) 10.0 psid (6)(7) 21.0 kW/ft (5) 1.30 (5)

<<10.6'./min of RATED THERMAL POWER (4'110.0>>

of RAT THERMAL POWER

<<9.7>> of RATED THERMAL POWER ALLOWABLE VALUES

<<2388 psia

~ 1821 psia (2) 43.7>>o (4) 91.5>>o (9)

~ 911 psia (3)

<<3.2 psig

<<0.118 psi/sec (6) (7)

~ 11.7 psid(6) (7)

<<10.2 psid (6)(7) 21.0 kW/ft (5) 1.30 (5)

<<11.0'/min of RATED THERMAL POWER (8)

<<111.0>> of RATED THERMAL POWER (8)

<<9.9% of RATED THERMAL POWER (8)

FUNCTIONAL UNIT REACTOR PROTECTIVE, INSTRUMENTATION'TRIP SETPOINT LIMITS TRIP SETPOINT ALLOMABLE VALUES m

2.

Logarithmic Power Level High (I) a.

Startup and Operating b.

,Shutdown C.

Core Protection. Calculator System 1

rcA r 1 '

1 A ~

L ctl

. lid I Cu Idloi S 2.

Core Protection Calculators D.

Supplementaiy Proieciion System Pressurizer, Pressure High II+

.RPS. LOGIC A.

Matrix. Logi c B.

Iniiiaiion Logic III. RPS ACTUATION'EVICES A.

Reactor Trip Breakers,.

B. Hanual'rip

.< 0.010% of RATED THERMAL POWER

~ 0.'010~ of RATED 1'IIrnAAAt nnurn v ncnIUnI ruilw Not Applicaole Not.Appli'cable:

~ '2409 psia Not Applicable Not Applicable Not Applicable.

Nnt Anpl 1 rabl e

. 0.011~, of RATED THERMAl POWER s 0.011~ of RATED

-THERYIAL POWER Not Applicable Not Applicable s 2414* psia Nnt Applicable

'Not Applicable Not Applicable Po+

App1 4 caL1 ~

'I C7

THIS PAGE INTENTIONALLYLEFT'LANK.

PALO VERDE UNIT 2 3/4 3-13 AMENDMENT NO.

19

TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREHENTS FUNCTIONAL UNIT I.

TRIP GENERATION A.

Process 1.

Pressurizer Pressure High CHANNEL CHECK CHANNEL CALIBRATION CHANNEL NODES IN WHICH FUNCTIONAL SURVEILLANCE TEST REIEEIEEE 1,

2 h

E

~ ~ l >>

a hva>>E

~

~ Eao>>

I nu co rr Yvvul I cYI rI Y~~ul Y = s.vn 3.

Steam Generator Level Low l

Cl l

I 1

Ul l

'1 ~

e3 CYIIIII IlYIIYIIllul LYVYI II I 'JJII 5.

Steam Generator Pressure Low 6.

Conthinment Pressure High 7.

Reactor.

Coolant Flow Low 8.

Local Power Density-- High-9.

DNBR Low B.

Excore Neutron Flux b

D(2,4),R(4,5j D

2, 4),

R -(4, 5) 8),

S (7) n n

Q, R(6j Q,

R (6) 1il C

1, 2

ll C

1, 2, 3*, 4*

1, 2

1.

2 1 ~

Variable Overpower Trip h

fEl l%

ll f 1 l\\

Q O (4) 2.

Logarithmic Power Level ¹igh C.

Core Protection Calculator System 1.

CEA Calculators R (4)

R Q and S/U (1) 1, 2, 3, 4, 5

(

and

Q,R(6) 1,2 2

Coi B Plotection Ca]culators D

2. 4). R(4,5)

Q(9), R(6) 1,2 V 8~,S(7)

TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE RE VIREO D.

Supplementary Protection System Pressurizer Pressure High II.

RPS LOGIC A.

Matrix Logic B.

Initiation Logic I I I.

RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manual Trip N.A.

N.A.

N;A.

N.A.

M, R(10) 1, 2

3*

4*

5*

I 2 3%'*

5*

1, 2, 3*, 4*, 5*

1 2

3*

4*

5*

I ~

C)

(2)

(4)

(5)

(7)

,(10)

'TABLE 4.:3-I~ContinueiQ REACTOR PROiTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS

'ABLE NOTATIONS

'With reactor trip. breakers in the closed position and the CEA drive system capablle of CEA withdrawal, and fue11 in the reactor vessel.

Eaich STARTUP or whien required with the reactor trip breakers c'losed anted the.CEA drive system capable of rod withidrawal, if not perf'or'med in the previous 7 days.

Heat balance on'ly (CHANNEL FUNCTIONAI'EST niot included),

above 15X of, RATED THERMAL POWER; adjust phd line'ar power 1@vel, thie CPC delta T

power ahd CPC nuclear power signal's to agree with the calbrimetric calculation if absolute differencd i< g'reste) than 2X. '.Dbring PHYSICS

TESTS, these daily calibraticins may.tie suspended provided these ca'librations are perf'ormed upon 'reaching;each major test power plateau and prior to proceeding to the next major test power platha0.

Above 15X of RAi'ED THERMAL POWER, 'verify that, the lihear power sub-channel gains, of the excore detector<

are consistent with the values used to establish the shape NnnLal'inq matrix elements in the Core Protection Calculators.

Neutron detectors ieay be-excluded fr6m,.CHANNEL CALIBRATION.

After each fuel'oading and prior to exceedir>g 70X of RATED THERMAL POWER, the incore detectors shall be usi d to determine thk saba)e annealing matrix e'lements and the Core, protection Calculators shall use these elements.

This iCHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel's close to the seniors ks Jralctica'ble to verif'y t)PERABILITY including alarm and/or trip functions.

Above 70X of RATED THERMAL POWER, verify that the total steady-'s'tate RCS flow rate as.indicated by each CPC is. less than or equal to the actual RCS total f'law raite determined by either using the reactor coolant pump differential presser'e in'strum'entat'ion or by-calorimetric calculations and if necessary, adjust.the CPC addressable constant flow coefficients such that e'ach CPC 'indic'ated flow is levels than 6r

'qual to the actual flow rate.

Thj flow measurement uncertainty may be included in the BEIRRl term ih the CPC ahd 'is equal to dr Qr6at5r

'hain 4X.

Above 70X of RATED THERMAL POWER, verify that the total sleazy-'state'CS flow rate as indicated by each CPC is 'less than or equal to the actual RCS total flow rate deterjmihed by e'ither using the

'react'or'oolant pump dif'ferential pressure instrumentation and th6 ultr~asbnic flow meter adjusted pump curves or calorimetric calculations The quarterly CtIAIIIIEL FUNCTIOliIAL Tl.:ST shal'I incllude verification that the correct (current) values of addressable constants are installed in each OPERABLE CPC.

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

PALO VERDE - UNIT',2 3/4:3-16 AMENDMENT NO.-W 64

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP, VALUES I

CD ESFA SYSTEM FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES B.

ESFA System Logic C.

Actuation Systems II.

CONTAINMENT ISOLATION (CIAS)

A.

Sensor/Trip Units 1.

Contaihment Pressure 2.

Pressurizer Pressure B.

ESFA System Logic C.

Actuation Systems III. CONTAINMENT SPRAY (CSAS)

A.

Sensor/Trip Units Containment Pressure B.

ESFA System Logic C.

Actuation Systems High Low High High I.

SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units 1.

Containment Pressure High 2.

Pressurizer Pressure Low 3.0 psig 1837 psia"'ot Applicable Not Applicable 3.0 psig 1837 psi a"'ot Applicable Not Applicable s 8.5 psig Not Applicable Not Applicable s 3.2 psig

~ 1821 psia"'ot Applicable Not Applicable s 3.2 psig

~ 1821 psia"~

Not Applicable Not Applicable s 8.9 psig Not Applicable Not Applicable m

CD rn 8

IV.

MAIN STEAM LINE ISOLATION (MSIS)

A.

Sensor/Trip Units 1.

Steam Generator Pressure Low 2.

Steam Generator Level High 3,

Containment Pressure - High B.

ESFA System Logic C.

Actuation Systems

= 919 psla(3) 91 Oro NR3 s 3,0 PSlg Not Applicable Not Applicable

~ 911 psla"'1.5%

NR"'

3.2 psig Not Applicable Not Applicable

TABLE 3.3-4 (Continued)

ENGINEEREO-SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES m

I C

ESFA SYSTEM FUNCTIONAL UNIT V.

RECIRCULATION (RAS)

A.

'Sensor/Trip Units Refueling Mater Storage Tank - Low B.

ESFA System Logic C.

Actuation Syste'a VI.

AUXILIARYFEEOMATER (SG"1)(AFAS-1)

A.

Sensor/Trip Units TRIP. VALUES..

7.4X of Span Not Applicable Not Applicable ALLOMABLE VALUES 7.9

> X of 'Span

> 6.9 Not Applicable Hot Applicable C ~

C'fnnH VA41I Stedm SG2 Lnnn>>ntn>> 41 I

a 1, I l1'41 ~ %

~ UlVI FJ f 4T4 I LVI I

Generator h Pressure-

.nc~.

(4)

CJ ~ QA Iln 185 psld (4)

CQ. JA 'WIC

< 192 psid m

m O

g)

~

B.

ESFA System Logic C.

Actuation SysteIIs VII. AUXILIARYFEEOMATTR $56-2)QFAS-2)

A Sensor/Trip Units 1.

Steam Generator f2 Level - Low 2.

Steaa Generatoi h Pressure-cn1

> cI c B.

ESFA Systea Logic C.

Actuation Systems VIII.

LOSS OF POMER A.

4. 16 kV Emergency Bus Undervoltage (Loss of Voltdge)

B.

4.16 kV Emergency Bus Undervoltage fnnn>>W W u (VLQIQllCU WV ~ E OLJC/

IX.

CONTROL ROOM ESSENTIAL FILTRATION Hot Applicable Hot Applicable

> 25.8X MR'43

< 185 psid Hot Applicable Hot Applicable

>,3250 'vo Its 2930 to 3744 volts with a 35-second maximum twmn dnlau

<,2 x 10-s pCi/cc Hot Applicable Hot Applicable

> 25.% MR/41

< 192 psid Hot Applicable Hot Applicable

> 3250 volts 2930 to 3744 volts with a-f5=second anvissum

+ ime del ~au

< 2 x 10-s pCi/cc

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS SFA SYST F

CTIONAL UNIT I.

SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units 1,

Containment Pressure

- High 2.

Pressurizer Pressure

- Low B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logic 3.

Manual SIAS C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays N.A N.A.

N.A.

N;A, N.A.

N.A.

CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST 0(2)

M(1) (3)

MODES FOR MHICH SURVEILLANCE III 1, 2, 3, 4

1, 2, 3, 4

II.

CONTAINMENT ISOLATION (C IAS)

A.

Sensor/Trip Units 1.

Containment Pressure

- High 2.

Pressurizer Pressure Low B.

ESFA System Logic 1,

Matrix Logic 2.

Initiation Logic 3.

Manual CIAS 4.

Manual SIAS N.A.

N.A.

1, 2, 3

I, 2, 3, 4

1, 2; 3, 4

1, 2, 3, 4

TABLE 4.3-2 -(Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION'SURVEILLANCE'RE UIREMENTS ESFA SYSTEM FUNCTIONAL UNIT CHANNEL CHANNEL'HECK CAL'IBRATION CHANNEL MODES FOR MHICH FUNCTIONAL SURVEILLANCE TE T:

IE REIEEIRER II.

CONTAINMENT ISOLATION (Continued)

C.

Automatic -Actuation Logic

'(exce'pt subgroup relays)

Act'uation,Subgroup Relays N.A.

N;A; N.A.

N.A.

~(2)

M(i) (3) 1', 2, 3, 4

.1, 2, 3, 4

Iii CONTAINMENT 5PRAY.(rSAS>

h C~aa ave/TE 4>> Ii~4 4~

fl~

sPLIIPMI ( I I 'I P,VII I Ir>

1'.

Co'ntainment 'Pressure, High -'igh R

1,,2, 3

B.

ESFA System Logic 1,

Matrix Logic 2.

-Initiati on-Logi-c 3 ~

Manual'CSAS N.A.

ll A

IIif\\~

N.A.

W A

llotl ~

0 1'23 a

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays N.A.

N.A.

~(2)

.W(I) (3) 1, 2, 3; 4

1,'2,3,4 m

CX

TABLE 4.3-2 (Continued)

ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS ESFA SYSTEM-FUNCTIONAL UNIT IV.

MAIN STEAM LINE ISOLATION (MSIS)

A.

Sensor/Trip Units 1.

Steam Generator Pressure Low 2.

Steam Generator Level High 3.

Containment Pressure - High CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST MODES FOR WHICH SURVEILLANCE 1,

2, 3, 4

1, 2, 3, 4

I, 2, 3, 4

B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logi'c 3.

Manual MSIS N.A.

N.A.

'.A.

N.A.

1, 2, 3, 4

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays N.A.

N.A.

0(2)

M(1) (3) 1, 2, 3, 4

1, 2, 3, 4

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES'CTUATION SYSTEM INSTRUMENTATION SURVEILLANCE'RE UIREMENTS ESFA SYSTEM. FUNCTIONAL UNIT V.

RECIRCULATION (RAS)

A.

Sensor/Trip Units Refuelinq Water Storage Tank Low 0

SCCA Ca ~ ~4 tata I tarsi t D ~

Cecal tl JJ> 44lll LV'JJ I I CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST MODES FOR WHICH SURVEILLANCE IS RE UIRED 1, 2, 3

1.

Matrix Logic 2i Initiation Logic 3.

Manual RAS C.

Automatic Actuation Logic (except s'ubgrnup re1ays)

Actuation Subgroup Relays VI.

AUXILIARYFEEDWATER (SG-ij (AFAS-ij A.

Sensor/Trip Units 1.

Steam Generator ffl Level-Low 2.

Steam Generator h.Pressure SG2 > SG1 N.A.

N.A'.

N.A.

N A N.A.

N.A.

N A N.A.

R, n(21 H(1] (3) 1,2,3,.4 1, 2, 3, 4

12, 3, 4

1, 2, 3

4 1, 2, 3, 4

1, 2, 3

~

(

C7m TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTBl INSTRUMENTATION SURVEILLANCE RE UIRENENTS CHANNEL NODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT ECK CRIER TI TE IE REIEUIREO VI.

AUXILIARYFEEDWATER (SG-1) (AFAS-1) (Continued)

B.

ESFA System Logic m

Cl m

CD

1. Hatrix Logic

2. Initiation Logic, 3.

Hanual AFAS C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays VII. AUXILIARYFEEDWATER (SG-2)(AFAS-2)

A.

Sensor/Trip Units 1.

Steam Generator 82 Level-Low 2.

Steam Generator A Pressure SGI ) SG2 B.

ESFA System Logic

1. matrix Logic

2. Initiation Logic 3.

Hanual AFAS C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays VIII. LOSS OF POWER (LOV)

A.

4.16 kV Emergency Bus Under-voltage (Loss of Voltage)

B.

4.16 kV Emergency Bus Under-voltage (Degraded-Voltage)

N.A.

N.A; N.A.

N.A.

O(2)

H 1

(3)

~ S2)

H 1)

(3) 1, 2, 3, 4

1, 2, 3, 4

1, 2; 3, 4

1, 2, 3, 4

1, 2, 3

1, 2, 3, 4

1, 2, 3,-

4 1, 2, 3, 4

1, 2, 3, 4

TABLk 4..3-2 (Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM I%'WJRVITKTT0W.WRVElTEUKElKgoTMERIW'3 TABLE N01'ATION (1)

Each train or logic channel shall be tj.sted at least every 62 days on a STAGGERI=D TEST BASIS.

(2)

Testing of automatic actuation logic shall include energization/

deenerqization of each initiation rely and verification of proper oper'at>on of each initiat;ion relay.

(3)

A subgroup,relay test shall ibe performed which shall include the energization/deenergization of each subgroup relay and verification'f the OPERABILITY of'ach subgroup re'lay'.

'Relays fisted below are exempt from testing during POWER OPERATION but shall be tested at least. ence per

'.18 months (luring REFUELING,'and during each COLD SHUTOOWN condition unless tested within the previous 62 days.

ACTUATION DEVICES THAT CANNOT -BE TESTED AT POWER TRAIN A ESF FUNCTION ACTUATION DEVICE:

TRAIN B ESF ACTUATION FUNCTION DEVICE SIAS A SIAS A CIAS A CIAS A CSAS A

HSIS A

AFAS 1A AFAS 2A K108 K409 K202 K204 K304 K305 K404 K211 K112 SIASIAS B,

CIAS B

CSAS B

HSIS B

MSIS B

AFAS 18 AFAS 18 AFAS 2B

', Kl,08

'4'09

'204 K304 K305 K404 K113 K211 K112 In the case of the following relays which are tested durin ower o)eration, one or more pieces of'quipment cannoWSe actuate,

u. can e rac e

ou bypassed or etc., which will 'not preclude the relay from being tested but will not actuate tllie locked out equipment associated'ith'h'e relay:

SIAS A SIAS A SIAS A CIAS A CIAS A RAS A RAS A RAS A AFAS 1A K401

K410 K412 K203 K210 K104 K312 K405 K113 SIAS 8 SIAS B

CIAS B

CIAS B RAS B

'RAS E3 RAS B

K301 K308 K203 K210 K104 K312 K405 PALO VERDE - UNIT 2 3/4 3-36

3 4.3 INSTRUMENTA BASES 3 4.3.1 and 3 4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES CTUATION SYST M INSTRUMENTATION The OPERABILITY of the 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 will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

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

The integrated operation of each of these systemsi 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.

Ol.

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 factors and CEA deviation penalties.

Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3. 1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.

The 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 AMENDMENT NO.~

64

INSTRUMENTATION BASES' TiTI<<l", OEilEIlD!

E~TF T,AT 5

E ontsnued The, measurettlent of iiesponse ti'me.at.the specified frequenciesprovides assur. ance that ttie'rotectiy'e and ESF, action function associated with each

.cha~nel's completed within tt>e tiaie limit,assumed in the safety "analyses.

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

The response'times in Table 3.3-2'are made up of the time to generate. the trip signal at thA d'etbct'or '(sens'or response time) and.

the time for the signal to interrupt power to'=the CEA drive mechanism (signal or trip delay time)..

Response

tittIe-roay be demonstrated

-by Any'6ries 'of'equential,, overlapping,

-oi total channel test measurements provided that such tests demonstrate the total channel response ti'me 'as defined.

Sdns'or,response time veri,fication may be demonstrated by either (1) in place, on~>it'e, 'or offsite test measurements'd (2) utilizing replacement sensors with certified riesponse,times.

PAlO VERDE - UNIT 2 B 3/4 3-2 AMENDMENT NO.

>9'

J v ~pea IIIQI c~

lgP

+

IA0 C'P UNITED'STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2O5S~I ARIZONA PUBLIC SERVICE COMPANY ET AL.

DOCKET NO.

STN 50-530 PALO VERDE NUCLEAR GENERATING STATION UNIT NO.

3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 License No.

NPF-74 1.

The Nuclear Regulatory Commissi'on (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 Authority dated August 5,

1993, as supplemented by letter dated January 19,

1994, complies wi,th 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; BE 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 act'ivities 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-74 is hereby amended to read as follows:

4k I

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 50, 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 Specifi'cations 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 and must be fully implemented no later than 90 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Mi

( ~ p~

Attachment:

Changes to the Technical Specifications Date of Issuance:

3uly 15, 1994 Theodore R. quay, Director

'Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

ii

~

L'

ATTACHMENT TO LICENSE AMENDMENT AMENDMB<T NO.

5O TO FACILITY'PERATING LICENSE NO.

NPF-74 DOCKET NO.

STN 50-530 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

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

Remove 2-3 3/4 3-14 3/4 3-15 3/4 3-16 3/4 3-23 3/4 3-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1 Insert 2-3 3/4 3-14 3/4 3-15 3/4 3'-16 3/4 3-25 3/4 3'-31 3/4 3-32 3/4 3-33 3/4 3-34 3/4 3-35 B 3/4 3-1

~

TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT 1.

TRIP GENERATION A.

Process 1.

Pressurizer Pressure High 2.

Pressurizer Pressure Low 3.

Steam Generator Level

Low 4.

Steam Generator Level High 5.

Steam Generator Pressure Low 6.

Containment Pressure High 7.

Reactor Coolant Flow Low a.

Rate b.

Floor c.

Band 8.

Local Power Density High 9.

DNBR Low B.

Excore Neutron Flux 1.

Variable Overpower Trip a.

Rate b.

Cei 1 ing c.

Band TRIP SETPOINT

<<2383 psia 1837 psia (2)

44. 2r, (4) 91.0~ (9)

~ 919 psia (3)

<<3.0 psig 0.115 psi/sec (6)(7) 11.9 psid (6)(7) 10.0 psid (6)(7) 21.0 I<W/ft (5) 1.30 (5)

<<10;6~/min of RATED THERMAL POWER (8)

<<110.0>> of RATED THERMAL POWER (8)

~9.7~ of RATED THERMAL POWER (8)

ALLOWABLE VALUES

<<2388 psi a

~ 1821 psia (2)

~ 43.7<o (4) 91.5-. (9)

~ 911 psia (3)

<<3.2 psig 0.118 psi/sec (6) (7)

~ 11.7 psid(6) (7)

<<10.2 psid (6)(7) 21.0 kW/ft (5) 1.30 (5)

<<11.0~/min of RATED THERMAL POWER (8)

<<111.0~4. of RATED THERMAL POWER (8)

~9.9~ of RATED THERMAL POWER (8)

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT 2.

Logarithmic Power Level - High (1) a.

Startup and Operating b.

Shutdown h

4 44 P

1

~ ~ 1 4

e' lr ~

'VVI K 6 I VlClr Lr IVII 40 I 'fall I dlVl 4+ 5 VVIII 1.

CEA Calculators 2.

Coie Protection Calculators TRIP SETPOINT

< 0.010X of RATED THERMAL, POwtR

< 0.010X of RATED THERMAL POWER Not Applicable Not Applicable ALLOWABLE VALUES

<'0.011X of RATED IHEKMAL POwER

< 0.011X of RATED THERMAL POWER Not Applicable Not Applicab(e D.

Supplementary Protection System rressurizer Pressure - High-

< 24% QSM 5 2414 psia II.

RPS LOGIC Matrix Logic B.

Initiation Logic iii. RPS ACTUATiON DEViCES

'A.

Reactor Trip Breakers B.

Manual Trip Not Applicable Not Applicable Not Applicable Not App'licable Not Applicable Not ADDlicable Not Applicable Not Applicable

'C3

THIS PAGE INTENTIONALLY DELETED PALO VERDE - UNIT 3 3/4'-13 PPENP~

NO.

18

TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIRENENTS FUNCTIONAL UNIT I.

TRIP GENERATION Ao PrUcess CHANNEL CHECK CHANNEL CALIBRATION CHANNEL NODES IN WHICH FUNCTIONAL SURVEILLANCE TERT

~RE ETRE 1.

Pressurizer Pressure - High 2.

Pressur'izer Pressure Low 3.

Steam Generator Level Low C+bhlll PNNNVlh+N'4 I NRRN1 U4NL hiIRuuiii uuiIuI uI uI I uVt I 5.

Steam Generator Pressure Low 6.

Containment Pressure - High 7

Roactnr Cnnlant Flnw I ow Local Power Density High 9l DNBR Low B.

Excore Neutron Flux 1

Variable Ovornnwor Trin K

Dn D (2, 4),

R (4, 5)

D (2, 4),

R (4, 5)

H (8),

S (7) h

~ I9 Ah M fO A%

ll (I,y 'TI g II ilI) 'Vl Q (4) 0 n

n 0,

R (6)

R (6) n l.

2 1

2 7

1 2

3*

4*

1, 2

lg C

1, 2

lly C

m C7 m

2 Logar'1 tiiigic 'Power Level Hlgh C.

Core Protection Calculator System 1.

CEA Ca1cul ators 2.

Core Protection Calculators R

Q (2, 4)I R (4, 5)

H (8),.

S (7) and S/U (1)

JI cy Dy 4I and

g, R (6) 1, 2

0 (9).,R (6)

Ii 2

I CD TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREHENTS

(

CD FUNCTIONAL UNIT D.

Supplementary Protection System Pressurizer Pressure High CHANNEL CHECK CHANNEL NODES IN MHICH CHANNEL FUNCTIONAL SURVEILLANCE LtB ATt II TEE

~EUIREO 1,

2 EA I

CJl II.

RPS LOGIC A.

Natrix Logic B.

Initiation Logic III.

RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Hanual Trip N.A.

N.A.

N, R(10) 3*

4*

5*

3%

4*

5*

3*

4*

5%'

3*

4*

5*

m CD

TABLE 4.3-3 +Continued' REACTOR PROTECTIVE INSTRUMENTATION SURVEI'LLANCE RE UIREMENTS (3)

(e)

(7)

(s)

'TABLE NOTA1 IOINS With reactor trip brealkers in the c'losed position and the OEA'r'ive system capable of CEA withdraw'al,'nd 'fuel in the reactor vessel.

Each SlARTUP.'r when required with the reactor trip breakers closed and the CIEA drive system capable of rod withdrawal, if not performed in the, previous

7. daiys'.

Heat bailance only (CHANNEL FUNCTIONAL TEST not included),

above 15-;;

of 'RATED THERMAL POWER", adjust the linear power level, the

~CPC delta T

power and CPC nuclear iIower signals to aglree with the calorimetric calculation if absolute differince is greater than 2~.

l)uring PHYSICS

TESTS, these daiily calibrations may.be suspended provided theRe calibrations are performed upoh

.r'caching 'each major test power plateau and prior to proceeding to the next major'6st power plateau.

Above l5~ of RATED THERMAL POWER, verify that the -linear power sub-channel gain,s of the excore 'detec'torls hre cons'istent w'ith the values used to establish the shape-anhea'lidg &at'rix elements in,the Corie Protection Calculators.

Neutron detectors may be.excluded. from CHANNEL CALIBRATION.

After each fuel loading and prior to exceeding 7G'f RATED THERMAL

POWER, the incore detectors shall:be used to determine the, shape annealling matrix elements and tIhe Core Prbtection Calculator's shall use. these elements.

This CHANNEL FUNCTIONAL TEST shal'I include the injection of simuIlated process signals into the channel as close to. the sensors as practicable to verify OPERABILITY includingj alar'm and/or trip functions.

Above 70$ of RATED THERMAL POWER, verify that the tcItal steady-state RCS f'low rate a,s indicated by each CPC is less than or equal to i;he actual RCS total flow rate determIined by either using the rhactorI coolant pump differential pressure ihstIrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that eac'h CPC'ndicated flow is less than or equal to the actual flow rate.

The flow measurement, uncertainty may be i'ncluded in the Bi.'RRl term in the CPC and is equal to or'great'.er'han 4~en Above 70~ of RAI'ED THERMAL POWER, verify that the total steady;-state RCS flow rate a.

indi catedl by.each CPC is 1'ess than or equal tb the~

actual RCS tota'I flow rate determined by either using the reactor coolant pump differential pressure instrumentation and the UltIIasonic flow imeter adju. ted pump curves or calorimetric calculations.

'he quarterly CHANNEL FUNCTIONAL TEST shall include verificaItiIrn that the correct (current) values of adldressable constants ar'e ihsth,lied in each OPERABLE'. CPC.

At least, once per 18 months and following maintenance or adj'ustme'nt'f thee reactor trip breakers, the CHANNEL FUNCTIONAL TEST shal'll include independen't verIificatiorr o'f the'ndervoltage and shunt'tr'ipse.

PALO VERDE " UN:[T 3 3/4 3-16 AMENDMENT NO M-c>P

TABLE 3.3-4 ENGINEER D SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.TRIP VALUES

(

m C7 m

C7 m

ESFA SYSTEM FUNCTIONAL UNIT I.

SAFETY INJECTION (SIAS)

A, Sensor/Trip Units 1.

Containment Pressure

- High 2.

Pressurizer Pressure

- Low B.

ESFA System Logic C.

Actuation Systems II.

CONTAINMENT ISOLATION (C IAS)

A.

Sensor/Trip Units 1.

Containment Pressure

- High 2.

Pressurizer Pressure

- Low B.

ESFA System Logic C.

Actuation Systems III. CONTAINMENT SPRAY (CSAS)

A, Sensor/Trip Units Containment Pressure High High B.

ESFA System Logic C.

Actuation Systems IV.

MAIN STEAM LINE ISOLATION (MSIS)

A.

Sensor/Trip, Units 1.

Steam Generator Pressure Low 2.

Steam Generator Level High 3.

Containment Pressure

- High B.

ESFA System Logic C,

Actuation Systems TRIP SETPOINT

< 3.0 psig

~ 1837 psia"'ot Applicable Not Applicable

~ 3.0 psig a 1837 psia"'ot Applicable Not Applicable s 8.5 psig Not Applicable Not Applicable

> 919 psia"'

91.0X NR'"

3.0 psig Not Applicable Not Applicable ALLOWABLE VALUES s 3.2 psig

> 1821 psia"'ot Applicable Not Applicable

~ 3.2 psig

> 1821 psia"'ot Applicable Not Applicable

< 8.9 psig Not Applicable Not Applicable

~ 911 psia"'

91.5/ NR"'

3.2 pslg Not Applicable Not Applicable

7.9

> X of Span

> 6.9 Not Applicable Not Applicable

> 25. 3g WR(

< 192 pseud

> 25,AX WR

< 185 psid Steam Genelator g Pressure-SG2

> SG1

't7 TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES PI ESFA SYSTEN FUNCTIOHAL UHIT TRIP VALUES ALLOWABLE VALUES v

RFrIRrUI ATION r RAS>

C.

d Cnrar~e If~4~ Ils4as n.

~ciiavi I

> Sy Villa' Refueling Natir Storage Tank -'ow 7.4X of Span B.

ESFA System Logic Not Applicable C.

Actuation System Hot Applicable VI.

AUXILIARYFEEINATER (SG-1)(AFAS-1)

A; Sensor/Trip Units 1.

Steam Generator ll Leve] - Low FSFA System Logic C.

Actuation Systems VII. AUXILIARY'FEEOWATER (SG-2) (AFAS-2)

A.

Sensor/Trip Units 1.

Steam Generator N2 Level - Low 2.

Steam Generator h Pressure-SG1 >

SG2 Noi Applicable Not Applicable

> 25.8X WR

< 185 osid Not Applicable Not Applicable

> 25. 3X WR( )

< 192 nsid B'.

y C.

'VIII.

A.

B.

ESFA SysteI Logic Actuation Systems LOSS Of POWER 4.16 kV Emergency Bus Undervoltage (Loss of Voltage)

4. 16 kV Emergency Bus Undervoltage (Begraded-Vol tage-)

Not Applicable'ot Applicable

> 3250 volts 2930 to 3744 volts with a 35-second maximum time delau Not Applicable Not Applicable

> 3250 volts 2930 to 3744 volts with-a 35-second iM~ ~xiii i. i iiii= uc i ay v4 ia 4 Can rln1 wi ~

IX.

CONTROL ROOM ESSENTIAL FILTRATION

< 2 x 10-s pCi/cc

TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE. RE UIREMENTS ESFA SYSTEM FUNCTIONAL UNIT I.

SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK CALIBRATION TEST MODES FOR MHICH SURVEILLANCE IS.RE UIRED 1.

Containment, Pressure High S

2.

Pressurizer Pressure Low S

B.

ESFA System Logic 1, 2, 3, 4

1, 2, 3, 4

1.

Matrix Logic 2.

Initiation Logic 3.

Manual SIAS N.A.

N.A.

¹A.

N.A.

1, 2, 3, 4

N.A.

2.

Pressurizer Pressure Low S

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays II. CONTAINMENT ISOLATION (CIAS)

A.

Sensor/Tri p Units 1.

Containment Pressure High S

N.A.

S(2)

M(1) (3) 1, 2, 3, 4

1, 2, 3, 4

1, 2, 3

B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logic 3.

Manual CIAS 4.

Manual SIAS N.A.

N.A.

V 1, 2, 3, 4

1, 2, 3, 4

TABLE 4.3-2 (Continued),

ENGINEERED SAFETY FEATURES ACTUATION.SYSTEM INSTRUMENTATION SURVEILLANCE. RE UIREMENTS ESFA,SYSTEM FUNCTIONAL UNIT II. CONTAINMENT ISOLATION (Continued)

,CHANNEL

.MODES FOR WHICH

'CHANNEL 'HANNEL FUNCTIONAL SURVEILLANCE CHECK CEIIEEKTICE

.. TE T

I

.REIIEI TEE C.

'Automatic. Actuation Logic

-(except. subgroup relays)

Actuation Subgroup Relays:

'N.A.

AI.A I) AHA N;A.

AI A

A+Ho Q(2)

M(I) (3) 1; 2, 3, 4

IA 2~. 3~, 4 III.

CONTAINMENT'.,SPRAY'CSAS) h ChtlC'AVIITVI46 IfhlO'K ni

~~i<av> I i s y v<>

< vv I1..

Containment

-'Pressure High - High

1,2,3

.B.

ESFA,System Logic

.I.:Matrix Logic, 2.:Initiation'ogic 3A-Manual.

CSAS N A:.

N.A'.

N.A.

,N.A.

I', 2, 3, 4

'1; '2, 3, 4.',2,3,.4 C.

EAuiomatic Actuation -Logic

.(except.subgroup relays)

Actuation Subgroup Relays N.A.

N.A.

'N.A.,

N.A.,

Q(2)

M(1) (3).

1, 2, 3, 4

1. 2, 3, 4

lD

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.SURVEILLANCE RE UIREMENTS

(

m Kl CD Pl ESFA SYSTEM FUNCTIONAL UNIT IV. MAIN STEAM LINE ISOLATION (MSIS)

A.

Sensor/Trip Units CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS RE(EUCHRED 1.

Steam Generator Pressure-Low S

2.

Steam Generator Level High S

3.

Containment Pressure -'igh S

1, 2, 3, 4

B.

ESFA System Logic 1.

Matrix Logic 2.

Initiation Logic 3.

Manual MSIS N.A.

N.A.

N. A-.

N.A.

1,2,3,4 1,

2, 3, 4

1, 2, 3, 4

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays N.A.

N.A.

0(2)

M(1) (3) 1, 2, 3, 4

m CD m

CD Ol CD

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS ESFA SYSTEM FUNCTIONAL UNIT V.

RECIRCULATION (RAS)

A.

Sensor/Trip.Units 0

C

~

1 I.I 4.

t L Rt Itt I I llg. ACltI Bl Vl d'LJV Tank - Low B.

ESFA'ystem Logic CHANNEL CHANNEL CHANNEL FUNCTIONAL CHECK CALIBRATION TEST MODES FOR WHICH SURVEILLANCE IS RE UIRED '

lp cy J

l Matrix I oriir 2.

Initiation Logic 3.

Manual RAS Automatic Actuation Logic

---(wxcept subgroup relaysj AcbLation Subgroup Relays Vl. AUXILIARYFEEDWATER (SG-I) (AFAS-I)

Seilsor/Trip Units 1.

Steam Generator ¹I Level-Lox 2.

Steam Generator z Pressure SG2

> SG1 KI h

I1 ~ ~ ~

N.A.

'A.H ~

N n

kl A

II ~ t1 ~

N.A.

w n

~ ~ ~ ~ I ~

Cl(2j L4/ 1 4 I'0 b

K~I K~I A

A L~

C~ S, 1,'2,3,4 I, 2, 3, 4

1, 2, 3, 4

1 9

'3 A

Cp 4) 't 1, 2, 3

1, 2, 3

C)

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREHENTS ESFA SYSTEM FUNCTIONAL UNIT CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST NODES FOR WHICH SURVEILLANCE IS RE UIRED VI.

AUXILIARYFEEDWATER (SG-1) (AFAS-1) (Continued)

B.

ESFA System Logic 1.

Hatrix Logic 2.

Initiation Logic 3.

Manual AFAS C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays VII. AUXILIARYFEEDWATER (SG-2)(AFAS-2)

A.

Sensor/Trip Units 1.

Steam Generator 82 Level Low 2.

Steam Generator h Pressure SGl ) SG2 B.

ESFA System Logic N.A.

N.A.

O2)

V i) (3) 1, 2, 3, 4

1, 2, 3, 4

I, 2, 3, 4

1, 2, 3, 4

I, 2, 3

1, 2, 3

1.

matrix Logic 2.

Initiati on Logi c 3.

manual AFAS N.A.

N.A.

1, 2, 3, 4

C.

Automatic Actuation Logic (except subgroup relays)

Actuation Subgroup Relays VIII. LOSS OF POWER (LOV)

A.

4. 16 kV Emergency Bus Under-voltage (Loss of Voltage)

B.

4. 16 kV Emergency Bus Under-voltage (Degraded Voltage)

N.A.

O2)

V 1)

(3) 1, 2, 3, 4

1, 2, 3, 4

TABLE 4. 3-2 ~Continued)

ENGINEERED. SAFETY FEATURES AC'IIUATION SYSTEM INSSSlUMI. STATEN '%%VEKCQZZ K U 1

S TABLE NOTATION (1)

Each-train or logic channel shal'1 be tes'ted at least every 62 da'ys'n a

STAGGERED TEST BASIS.

(2)

Testing of automatic actuation logic shall include energization/

deenergization of each initiation relay and verification of proper operation of each initiation relay.

(3)

A subgroup relay test shall be performed whicfi s'ha11 include the energization/deenergization.

of ekch sL>bgroup relay 'and verification of the OPERABILITY of each subgroup relay.

Relays listed below are exempt from testing during POWER OPERATION but shall be tested. at least once per 18 months during REFUELING and during each COLA SHUTDOWN condition unless tested'withi'n the'r'evious 62 days.

ACTUATION DEVICES TNAT, CANNOT BE 'TESTED AT P(NER TRAIN A ESF FUNCTION SIAS A SIAS A

CIAS A C!AS A CSAS A

HSI 5 A

HSIS A

AFAS 1A AFAS 2A ACTUATION DEVICE KI.08 K409 K202 K204 K304 K305 K404 K211 K112 TRAIN '8 LSF FUNCTION SIAS 8 SIAS 8 CIAS 8 CSAS 8

NSIS 8

HSIS 8 AFAS'8 AFAS 18 AFAS 28 ACTUATI'ON DEVICE K108 K409 K204 K304 K305 K404 K113 K211 K1.12 In the case. of the fol,lowing relays-wh'ich, one or more pieces of 'equipsient cannot be bypassed or etc., which will not preclude will not actuate the locked out equipment are tested durin ower o

erhti'on'ctuated, but Can be racked out, the -relay from being tested but associated with the relay:

SIAS A SIAS A SIAS A CIAS A CIAS A RAS A

RAS A RAS A AFAS 1A K401 K410

'K412 K203 K210 K104 K312 K405

K113, SIAS 8 SIAS 8

'CIAS 8 CIAS 8 RAS 8 RAS 8 RAS 8 K301 K308 K203 K210 K104 K312 K405 PALO VERDE - UNIT 3 3/4 3-36

~ ~

3 4.3 INSTRUMENTAT BASES 3 4.3.1 and 3 4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the reactor protective and Engineered Safety Features Actuation Systems instrumentation and bypasses ensures that (1) the associated Engineered Safety Features Actuation action, and/or.reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliabil,ity, 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 CEJ; "RPS/ESFAS Extended Test Interval Evaluation,"

and CEN-327-A, Supplement 'Ip 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 factors and CEA deviation penalties.

Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3. 1 and 6,8.1.)

ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.

The 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 3 B 3/4 3-1 AMENDMENT NO- +8-s

INSTRUMENTATION BASES-REACTOR'PROTECTIVE AND I:NGINE~RIP!SAFETY F.ATUR S

CTUATION SYSTEN INSTRUMENTATION (Continued)

The measurement of response time. ath the specified fr e'qu'enci'es.

ptiovides assurance tha1t; the protective an@'.,ESF action functio'ri associatedwith

'each chan'nel is', comp'leted wit;hin the time limit aslsumed in"the safety ana'lyses.,

No'dcredit

'was 'takerid in the anIalyse.'3'or 'thyrse channels with response timeS

'indicated as not applicablie.

Ttre response times in Table 3.3-2.are made up of the time to generate the t'ripI:signhrl.at Itha detectrrr '(sensor resp'onse time) and the time for the s'ignal to initervupt. pokerlto the.CEA drive mechanism (signal or trip delay. time).

Response"time may be dern'onstrated by any'drihes of sequential, rrverlapping, or total charm'el test measur'emer!ts provided that such"tests demhonstrate the total

.channel response t;ime as dhefined.

'Sensor res'ponse time verification may be demonstrated by-either:(1) in pllace,.onSite, 'or'offsi'te'test 'measurerrrents,or (2) utilizirig replacement

.serisors with dertified response times.

3 4:3';3.1 RAI)IATION MOhliTIGRING INiiTRUMENTATION The OPERABILI;llY of thle. radiiatiori mdnitorin'q channels ensures thIrt:

(1);the.radiat'ion,)level's, ar'e contir>uallj'easured in the areas served. by the

'individual channel'..

and '(2) the alarm ort'utomatic. act'idn is initiated when the radiation 1evell trip shetpoitrt 'is exceeded.

The. OPERABILITY'bf. thhe incore detedtoiI;s with the specified minirrIurtr comple~

me'nt of equipment, ensures that tjr'e 'measure'itients'obtained from:use. of this system accurately drepresent the slpatia1I'neutror! flux distrib'ution of the reactor Icore.

3'.3.3.3 SEISMIC INSTJ'lUHIENTATl[N The OPERABILITY of thle. sei!vie instIrutlrentation ensures

'that sufflicient capability, is availlable to promptly determine the, magnitude of a,seismi'c,everit and evaluate.the response lof those featured ijtrplrtant to safety.

This capability is required to permit, comparison of.the measurecl riespon! e to that used lin the design basis for the facility to determine if plant shutdown -is, requ'ired,pursuant to Appendix A of 10 CFR Pa'rt 100.

The 'instrumentatiorI is consistent wilth thle

',recommenda'tio!~s of Regul athory Grride 1.12, "In'strumentation for Earthquake~"'pril 1974 as iden'ti'fi'eel iIo the PVNGS"FSAR,'.

Th(I seismic instrumentation for the.

site is listed'n Table 3.3-7.

d The'OPERABILI'll'Y 'of thie metcerologidal 'instr'umentation. ensures that suffi-

.cient meteoro',logical data are avai'liable Ifor estimating potential radiatiion ddses to the public as a result Iof,ror!tine or 'acilid'ental release of radioactive PALO VERDE UNIT 3 B 3/4 3-2.

AMENDMENT NO.

18

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