Text

Amends 98,86 & 69 to Licenses NPF-41,NPF-51 & NPF-74, Respectively,Clarifying Shutdown Margin Definition,Changing Shutdown Margin Applicability & Surveillance Requirements to Comply W/Safety Analysis Assumptions
	ML17311B166
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Site:	Palo Verde
Issue date:	09/01/1995
From:	Thomas C; NRC (Affiliation Not Assigned)
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UNITED:STATES NUCLEAR.REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 R

ONA U

CSRVC COMPAY TA.

.'CK T NO.

STN 50-5 8

PA 0 VERD NUCL AR G

N RATING STATION UNIT NO.

1 NDM NT TO FAC ITY OP RAT NG C

NS Amendment No.

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 California Edison

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

Power, and Southern California Public Power Authority dated March 31,

1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance

(:i) that the activities authorized by this amendment can be conducted without endangering the health 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 heal.th and safety, of the public; cLAd 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'o read as follows:

9SO9OSOS2e 9sO90X PDR ADGCX 0SaOOS28 P

PDR

4 C'

~ )

' 2) echnical S ecifications and Environmental Protection-Plan The Technical Specifications contained in Appendix A', as revised through Amendment No.

9B, 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.

The license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION I

Charles R. Th'omas, Project Hanager

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

Attachment:

Changes to the Technical Specifications Date of Issuance:

SePtember 1,

1995

41 I

. ~

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~1

ATTACHMENT TO LICENSE AMENDMENT M NDM T NO'.

98 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.

The corresponding overleaf pages are also provided to maintain document compl'eteness.

REMOV III*

IV 1-6 3/4 1-1 3/4 1-2 3/4 1-3 3/4 3-3 3 7*

3/4 3-8 3/4 3-16 3/4 9-1 B 3/4 1-1 B 3/4 1-la B 3/4 9-1 B 3/4 9-2*

6-20a INSERT III*

IV 1-6 4/4 '1-1 3/4 1-2 3/4 1-3 3/4.3-3 3-7*

3/4 3-8,'/43-16 3/4 3-'16a, 3/4 9-1 B 3/4 1-1 B 3/4 1-la B 3/4'-1 B 3/4 9-2*

6-20a

No changes were made to these

pages, reissued to become overleaf -pages.

41 Q>

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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION'.1 SAFETY LIHITS PAGE 2.1.1 2.1.1.1 201.1.2

'2. l. 2 REACTOR CORE 0 ~ 0000 ~ 000 ~ ~ ~ 0 ~ ~ ~ ~ ~ ~

PEAK LINEAR HEAT RATE............

REACTOR COOLANT SYSTEM PRESSURE..

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ '

~ ~ ~ ~ ~ ~ ~ ~ '

~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ ~ ~ ~ ~ ~ ~ ~; ~ ~ ~ ~

~

~ ~ ~

~

~ ~ ~ ~ ~

2-1 2-1

'2-1 2-1 2.2 LIMITING SAFETY SYSTEMSETTINGS 2.2.1 REACTOR TRIP SETPOINTS...........................'............

2-2 BASES SECTION 2

1 SAFETY LIMITS

2. 1.1 REACTOR CORE................................-....-.......--

2. 1.2 REACTOR COOLANT SYSTEH PRESSURE............................

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETPOINTS.....................................

P.AGE B 2-1 8 2-2 B 2-2 PALO VERDE UNIT 1

T~ND X

LIMITING CONDITIONS FOR OPERATIOIN AND SURV ILLANCE RE U'(REHENTS SECTION E

TTI 'ETEI.

EIE 3/4. 1'. 1 BORATICIN CONTROL SHUTCIOWI,'l hIIARGIN - REACTOR TRIP BREAKERS OPEN..........

SHUTCIOWIN hIIARGIN REAC1'OR TRIP BREAKERS CLOSED........

MODEFNTOR TEHPERA'll'URE COEFF ICIENT'...'.'..........,......

MINIl'lUH TEMPERATURE FOR CRITICALITY...................

F'AGI(

3I/4 0-1 3/4 l-l 3I/4 1-2'/4 1-4 3/4 1-5 FLOW PATHS-OPERATING........,....

CHARG!ING PUHPS - SHUTDOWN..... I,...

CHARG!ING PUMPS - OPERAl ING.... 3...

BORAl'ED WATER SOUftCES - SHUTDOWN..

BORAl*ED WATER SOUf(CES - OPERATING.

BOROhl DILUITION ALARMS.........,.... ~

~ ~

~

~ ~

~

~ ~

~

~ ~ ~

~ ~ n 3/4.1. 2 BORATICIN SYSTEMS FLOW PATHS SHUTDOWN.........,......,................,

'I/4 1-6 3/4 1-7 3/4 1-8 3/4 1-9 3I/4 1-10 3/4 1-12 3/4" 1-13

~

~ ~

~

~ 0 ~

~ ~ ~

EE POSITIOIN INDICATOR CHAhlNELS '- SHUTDOWN...'. ~

0

~

~ ~ 0

~ ~ ~

~

EE CEA DROP TIMID....,...................,.................

~

SHUTDOWIN C',EA IN'SERTION LIMIT..............

REGULATINGI CIEA INSERTION LIMITS..'...'...'...',.

PART LENG1'H CEA INSERTION LIMITS......................

3/4.1.3 MOVABLE CIONTROI. ASSEMBLIES CEA POSITION......,........................

POSI ITIiON INDICATOR CHANNELS OPERATING.'.

'I/4 1-15 3/4 1-17 3/4 1 3/4 1-19 3/4 1-20 3/4 1-21 3/4 1-23 PALO VERDE UNIT 1 IV Amenclmeit No. &HA,98

DEFINITIONS R

PORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Sections 50.72 and 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

No change in part-length control element assembly position, and b.

All full-length control element assemblies (shutdown and'egulating) are fully inserted.except for the singl'e assembly of hi'ghest reactivity worth"which is assumed to be ful:ly. withdrawn.

With any full-length CEAs not capable of being fully inserted, the withdrawn reactivity worth of these full.-length CEAs must be accounted far in the determination of the SHUTDOWN'ARGIN.

SIT BOU DAR 1.30 The SITE BOUNDARY shall be that line beyond which the land, is neither

owned, nor leased, nor otherwise control-led by the licensee.

$0DT~WAR 1.31 The digital computer-SOFTWARE for the reactor protection system shall be the program codes including their associated data,. documentation, and procedures.

SOURCE CH CK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

S GG T BASIS 1.33 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains,. or other designated components obtained by.dividing the specified test interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERHAL POWER 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

PALO VERDE - UNIT 1 1-6

'Amendment No. Q3-;68,98

J I

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ACTIVITY CONTROL SYSTEMS 3 4. 1 REACT VITY CONTROL SYSTEMS 3 4.1.1 BORATION CONTROL S

UTDOWN MARGIN REACTOR TRIP BREAKERS OPEN**

LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0X delta k/k.

3>>LI LITT: Ilpddpp;I, dd It 3

ttpt I

3

~CT ON:

With the SHUTDOWN MARGIN less than 1.0X delta k/k, 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.

SURVEILLANCE RE UIREMENTS

4. 1. 1. 1. 1 The SHUTDOWN MARGIN shall be determi'ned to be greater than or equal to 1.0X delta k/k 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 bur nup based on gross thermal energy generation, Xenon concentration, and Samarium concentration.

4. 1. 1. 1.2 The overall core reactivity balance shal,l be compared to predicted values to demonstrate agreement within 2 1.0X delta k/k at least once per 31 Effective TFull '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 (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. 28, 98

SHUTDOWN MARGIN - REACTOR TRIP'REAKERS CLOSED**

LIMITING CONDITION f'R OPERATION 3.1.1.2 a.

The SHUTDOWN MARGIN shall be greaiteit'han

'or equal to that specified in the CORE OPERATINIG I IMITS REPORT,

and, b.

For T,~ less than or. equal to, 5)0.(=, K, sha(l be. less thlan 0.99.

c.

Reactor criticality shall not be aclhieved with shutdown grbuP CEA movement.

I>>L<<: >>:,, ',;'

closed.**

~CTI ON:

a ~

b.

With the SHUTDOWN MARGIN'ess than that speci. fied,in the CORE OPERATING LIMITS REPORT, immedi'ately 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 riequired SHUTDOWN MARGIN is restored, and With T,~ less than or equal to F00'F and K., greater than'r elqual to 0.95, immediately vary CEA positions and/or initiate and continue boration at greater than or equal to 26 gpm to the reactors coolant system of a solution containing greater, than or equal to 4000 ppm, boron or equivalent until the required K, is restored.

SURVEILLANCE RE UIREMEINTS

4. 1. 1.2.1 With the reactor trip breakers closed**,.the SHUTDOWN MARGIN shall be determined to be greater than or equal to that specified in the CORE OPERATING LIMITS REf'ORT:

a.

Within 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 wlhile the CEA(s) is inoperablh.

See Special Test Exceptions 3.10.1 and 3'.10.9,,

- The CEA drive system, capable of CEA withdrawal.

PALO VERDE - UNIT 1 3/4 1-2 Amendment No..83-,69,98

CTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREHENTS Con'tinued b.

When in MODE 1 or MODE 2 with k, greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3. 1.3.6.

If CEA group withdrawal is not within the Transient Insertion Limits of Specification 3. 1.3.6, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months verify that 'SHUTDOWN MARGIN is greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

c.

When in MODE '2 with k <> less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criC>cality by verifying that the predicted critical CEA position. is within, the limi,ts of Specification 3. 1.3.6.

d.

Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors, of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e.

When in MODE 3, 4, or 5, 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. l. 1.2.2 When in MODE 3, 4, or 5, with the reactor trip breakers closed**.

and T<~ less than or equal to 500'F, K<.

shall be determined to be less than 0.99 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by const'Jeration 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. 1.2.3 When in MODES 3, 4, or 5 with the reactor trip breakers closed**,

verify that critical.ity cannot be achieved with shutdown group CEA withdrawal 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. 1.2.4 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. 1. 1.2. l.e or 4. 1.1.2.2.

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.

- The CEA drive system capable of CEA withdrawal.

PALO'ERDE UNIT 1 3/4 1-3 Amendment No. 83-~,98

REACTIVITY CONTROI SYS'll EMS MODERATOR TEMPERATURE COEFFICI ENl LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coelfficient (MTC) shall be within the area of Acceptable Operation slpecified 'in'the CORE OPERATING LIMITS4REPORT.

The maximum. upper limit shall be less than or equal to + 0.5i x 10 hK/K"F for a power level of OX RATED TIHE$IMAL POWER'with a linear ramp to OBK/K/ F at lOOX RATED THERMAL POWER.

APPLICABILITY:

Mf)DES 1 and 2"8.

ACTION:

With the moderator temperature coefficient outside the area of Acceptable Operation, be in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;s.,

SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The INC;shall be determined,to be within its limits by c'onfirmatory measurements.

INTC measured va'ilues shall be extrapolated an'd/or compensated to permit direct compariso'n with the above limits.

4.1.1.3.2 The ICC sha'il be determined at, the following frequencies And'THERMAL'OWER conditions during-each fuel cycle:

a.

Prior to initia1 operation abov'e 5X 'of RATED THERMAL POWER,, after each fuel loading.

b.

.C.

At any THERMAL POWIER, within 7 EFPD after reaching a core, average exposure of'0 EFPD bturnup into the current cycle.

At any TiHEBQIL POWER, within 7 EFPD after reaching a core average exposure equivalent to two-thirds of the expected current cyc1le end-of-cycle core average burnup.

"WithKeff greater than or equal to 1.0.

O'See Special: Te t 'lException 3.10.2.

PALO VERDE " INIT 1 3/4 1-I4 AMENDMENT NO.g7,"

69

FUNCT ONAL UNIT 1.

TRIP GENERATION A.

Process TAB E.3.3-EACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP.

~OPERABL ROBES

)(~CION 2.

3.

5.

6.

7.

8.

9.

Pressurizer Pressure - High Pressurizer Pressure - Low Steam Generator Level - Low Steam Generator Level High Steam Generator Pressure Low Containment Pressure - High Reactor Coolant Flow - Low Local Power Density - High DNBR - Low 4

4 4/SG 4/SG 4/SG 4

4/SG 4

4 2

3 2 (b) 3 2/SG 3/SG 2/SG 3/SG 2/SG 3/SG 2

3 2/SG 3/SG 2 (c)(d) 3 2 (c)(d) 3 1,2 1,2 1,2 1,2 3* 4*

1,2 1,2 1,2 1,2 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ 2¹,3¹ B.

Excore Neutron Flux 1.

Variable Overpower Trip 2.

Logarithmic Power Level - High a.

Startup and Operating b.

Shutdown C.

Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators 2(a)(d) 2 0

1 2 (e) 2 (c)(d) 3 1,2 2¹,3¹ 1,2 2¹,3¹ 3* 4* 5*

9 3,4,5 4

1,2 6,7 le2 3* 4* 5* 25a35 7

10 I

Ik

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l k$

~ '

TABLE 3.3-1,(Continued)

ACTION STATEMENTS 3.

Steam Generator Pressure Steam Generator Pressure

- Low Low Steam Generator Level'-Low (ESF)

~ Steam Generator Level 2-Low (ESF) 4.

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

(Wide Range)

Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF) 5.

Core Protection Calculator.

Local'ower Density - High (RPS)

.DNBR - Low (RPS)

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

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

ACTION 4 With the number of channels OPERABL'E one less than required by the Minimum Channels OPERABL'E requirement, suspend all operations involving positive reacti.vity changes.

ACTION 5 With the number of channels OPERABLE one less than required by the Minimum Channels..OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breaker of the inoperable channel is, placed in the tripped'ondition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise, be.in;at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;

however, the trip breaker associated. with the inoperable channel may, be closed for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing, per Specification 4.3.1.1.

ACTiON 6 a.

With one CEAC inoperable, operati'on may con inue for up to 7 days provided that the requirements of.Specification 4. 1.3. 1. 1 are met.

After 7 days, operation may continue provided that the conditions of Action Item 6..b are met.

b.

With both CEACs inoperable, operation may continue provided that:

1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the,DNBR margin required 'by Specifica-

,tion 3. 2. 4. b (COLSS in service) or 3. 2.4. d (COLSS out of service) is satisfied, and the Reactor Power Cutback System is disabled, and PALO VERDE,UNIT1 3/4 3-7 AMENDMENT NO. 50

ACTION STAIENENIS ACTION 7 ACTION 8 2.

Wit,hin 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

All full-length and part-length CEA groups must

'be withdrawn withiii"the limits of Spiacifications 3.1.3.5, 3; 1.3.6b, and 3. 1.3.,7b except during

.surveillance testing pursuant to the requirements of Specification 4. 1.,3. 1.2.

.'Specification 3.1.3.6b allow..

CEA group 5

inser~tion to no further than 127.5 inches withdlrawn.

b)

'The "IRSP'T/CEAC Inoperable" addres. able cons<ant; in the CPCs is set to be indicated that both CEAC',s.are inoperable.

c)

The.Control Element, Drive Mechanism Controls Systeim (CEDMCS) is placed in and

.ubsequently maintained in the "Standby" mode except during CEA motion permitted by Specifications 3.1.3.5 3.1.3.6bI, and 31;3.7b when the CEDMCS may be, operatedl in either the "Manual Group" or "Manual Individual"'ode.

3.

CEA position sulrveillance must meet the requiremgntIs of Specifications

4. 1.3. 1. 1, 4.1.'3.5,

4. 1.3,6,I and I

4. 1.3.7 except during surveillance testing pursuant to Sp'ecification 4.1.,3.1.2.

With three or more auto restarts,.

excluding periodic auto restarts (Code 30 and Code 33) of one non-bypassed calculytop du'ring a 12-hour interval,, demonstrate calculator OPERABILITY by performing a CilANNEL FUNCTIONAL TES'1 iii,thin the next 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ',.

With the number of'PERABI E channels one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel to OPEfNBLI: status within 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . or open an affected reactor trip breaker within the next hour.

ACTION 9

With the number of OPERABLE channels,.

one less than the Mini'mum Channels OPERABLE r'equirement, 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 open the reactor trip breakers within the next hour.

ACTION 10 In MODE'S 3, 4, or 5, the Core Protection Ca'iculator channels are not required to be OPERABLE when the Logarithmic Power, Level High trip is OPERABLE with the trip setpoint lowered to gl0 X of Rated Thermal Power.

i PALO VERDE UNIT' 3/4 3-8 Amendment No. 56I,98

TABLE 4.3-.

(Continued)

REACTOR PROTECTIVE INSTRUHENTATION SURVEILLANCE RE UIREHENTS FUNCTI NA UNIT 0.

Supplementary Protection System Pressurizer Pressure

- High CHANNEL CHANNEL CHECK CALIBRATION CHANNEL NODES IN WHICH FUNCTIONAL SURVEILLANCE TEST.

~EEEEE 1,

2 II.

RPS LOGIC A.

Hatrix Logic B;

Initiation Logic III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Nanual Trip N.A.

N.A.

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

1 2

3*

4*

5*

1, 2, 3*, 4

5*

i, r, 3*, 4+, 5*

m C7 C)

TA~BE 4. 3 ~1'tintirl<ied)

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

(1)

Each STARTUP or when required wi,th,the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.

(2)

Heat balance only (CHANNEL FUNCTIONAL TEST not included):

a.

Between 15X and 80X of RATED THERhiAL POWER, compare the linear power level,, the CP~C delta T power and the CPC nuclear power signals to >the calorimetric calculation.

If any signal is within -0.5X to 10X of the calorimetric then do not cal'ibrate except as requi>red, during initial power ascension after reiFueling.

If any signal is less than the calorimetric calculation by more than 05X, then adjust the affected signal(s) to agree with the calorimetric calculation.

If any signal is greater than the calorimetric calculation by more than 10X then adjust the affected signal(s) to agree with the calorimetric calculation within 8X to 10X.

b.

At or above 80X of IRATED 'THERhQL POWEIR; compare the linear power level the CPC delta

~T power~and the CPC nuclear power signals to the caloirimetric calculation.

If any signal differs from the calorimetric calculation by an absolute difference of more than 2X, then adjust the affected signal(s) to agree with the calorimetric calculation.i During PHYSICS TESTS, these daily dalhbriations may be suspended provided these calilbrations are performed upon reaching each major test power plateau and prior to proceeding, to the next major: test, power plateau.

(3)

Above 15X o/F RATED THERhiAL POWER',, Verifp that the linear powers subchannel gains of the excore detectors are, consistent with the values used to esta'blish the shape annealing mat'rix elements in the Core Protection Calculators.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

After each IFuel lloading and prior to. exceeding 70X of RATE'.0 THERMAL POWER, the incore detectors shall be used to determine the shape annealing matrix elements and the Core Protection Calculators shall use these elements.

'PALO VERDE - UNIT 1

3/4 3-16 Amendment No. &7-,78, 98

TABLE 4.)-l Continued TABL'E NOTATIONS (6)

This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as close to the sensors as practicable to verify OPERABILITY including alarm and/or. trip functions.

(7)

(8)

Above 70X of RATED THEfNAL POWER, verify that the total steady-state RCS flow rate as indicated by each CPC is less than or equal to the actual.

RCS total flow rate determined by either using the.reactor coolant pump differential pressure instrumentation or by calorimetric calculations and.if necessary.,

adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty may be included in the BERRl team in the CPC and is. equal to or greater than 4X.

Above 7OX of RATED THERMAL POWER, verify that the total steady-state RCS.flow rate as. indicated by each.

CPC is less. than or equal to the actual RCS total flow rate determined, by either using the reactor coolant pump differential pressure instrumentation and the ultrasonic flow meter adjus'ted pump curves or calorimetric calculations.

(9)

The quarterly 'CHANNEL FUNCTIONAL TEST shall include verification that the correct current values of addressable constants are installed in each OPERABLE CPC.

(10)

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

PALO VERDE UNIT I 3/4 3-16a Amendment No. 87-,78,. 98

~l.

', ~

"a 4

3 4.9 REFUELING OPERATIONS 3 4.9. 1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9. 1 With the reactor vessel head closure 'bolts less than fully tensioned or with the head'emoved, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and within the limit specified in the Core Operating Limits Report (COLR).

~LI A T '.

IIOOE 6'.

~CT ON:

'With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than. or equal to 26 gpm of a solution containing

> 4000 ppm boron or its equivalent until the boron concentration is within l.imits.

SURVEILLANCE RE UIREMENTS 4.9. 1. 1 The boron concentration shall be determined to be within.the limit specified in the COLR prior to:

a.

Removing or unbolting the reactor vessel

head, and b.

Withdrawal of any full,-length CEA in excess of 3 feet from its fully

,inserted position within the reactor pressure vessel.

4.9. 1.2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined'y chemical analysis -at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.

PALO VERDE UNIT 1 3/4 9-1 Amendment No. 54,9Q

REFUELING OPERATIONS 3/4.9.2 INTRUHENTATION LIMITED CONDITION FOR OPERATION 3.9.2 As a minimum, two startup channel c>eutrdn flux monitors shall be OPERABLE and operating, each with continubus'isual indication in the control room and one with audib1le indication in the conta'inment and control room!,

APPLICABILITY:.MODE 6.

ACTION:

'a ~

b.

With one of the above required 6onhtdrs'n'operable or not operating, immediately suspend all operatidns inlvolving, CORE ALTERATIONS or positive reactivity changes.

'ith both of the above required 'monit'orS inoperable.or not ope0at$ ng, determine the boron concentration of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.9.2 Each startup channel neutron flux monitor shall be demonstrated OPERABLE by performance of:

a.

A CHANNEL CHECI( at 'least once per 12 t~ours, b.

A CHANNEL FUNCTIONAL TES1 within'8 'hours'rior to the initial start of CORE ALTERATIONSand c.

A CHANNEL FUNTIONAL TEST at least o'ncaa per 7 days.

PALO VERDE - UNIT 1 3/4 9-2 AyENOHIENT NO.

3 4-1 REACTIVITY CONTROL SYSTEMS BASES 3 4.1.1 BORATION CONTROL 3

. l.

. 1 d 3 4. 1. 1.2 SHUTDOWN MARGIN AND 'K The function of SHUTDOWN MARGIN is to ensure that,the reactor remains subcritical following a design basis accident or anticipated operational occurrence..

The function of K,, is to maintain sufficient subcri,ticality to preclude inadvertent criticali'ty following ejection of a single control element assembly (CEA).

During operation. in MODES 1 and 2, with k << greater than or equal to 1.0, the transient insertion limits of Speci'ficatlon 3. 1.3.6 ensure.that sufficient SHUTDOWN MARGIN is available.

SHUTDOWN MARGIN's the amount by which the core is subcritical, or would be subcritical immediately following a reactor trip, considering a single malfunction

. esulting in the highest worth CEA fail-ing to ins"~t.

K, is a

measure of the core's reactivity, considering a single mal.function. resulting in the highest worth inserted CEA being ejected.

SHUTDOWN MARGIN requirements vary throughout the core life as a function of, fuel depletion and reactor coolant. system (RCS) cold leg temperature (T<<)..

The most restrictive condition occurs at EOL, with T<< at no-load operating temperature, and is associated with a postulated, steam line break accident and the resulting uncontrolled, RCS cooldown.

In the analysis of this

.accident, the specified SHUTDOWN MARGIN is required'. to control the reactivity transient and ensure that the fuel,performance and offsite dose criteria are satisfied.

As (initial) T,<< decreases,,the potential RCS cooldown and the resulting. reactivity transient are less severe and, therefore, the required'HUTDOWN MARGIN also. decreases.

Below T<< of about 350 F, the inadvertent deboration event becomes 1:imiting.with respect to the.

SHUTDOWN MARGIN requirements.

Below 350'F, the specified. SHUTDOWN MARGIN, ensures that sufficient, time for operator actions exists between the initial indication of the deboration and the total, loss of shutdown margin.

A'ccordingly, with -the reactor trip breakers closed and the CEA drive system capable of CEA withdrawal, the SHUTDOWN MARGIN requirements are based.

upon these l.imi,ting, conditions.

Additional events considered

.in establishing requirements on SHUTDOWN MARGIN that are not limiting with respect to the Specification l.imits are single CEA withdrawal and startup of an. inactive reactor coolant pump.

K, requirements vary with the amount of positive reactivity that would be introduced assuming, the CEA with the highest inserted worth ejects from the core.

In the analysis of the CEA ejection event, the K requirement ensures that the radially averaged enthalpy acceptance criterion is satisfied, considering power redistribution effects.

Above T<< of 500 F, Doppler reactivity feedback is sufficient to preclude the need for a specific K, requirement.

With all CEAs fully inserted, K, and SHUTDOWN MARGIN requirements are equivalent in terms of minimum acceptable core boron concentration.

PALO VERDE UNIT 1 B 3/4 1-1 Amendment No. 83, 98

R ACTIVITY'-CONTROL SYSTEMS, BASES~N" 'i

.'( )" i"l d)

The requirement, prohibiting-criticality due. tci shutdown group'EA, movement is associated:with the as'sumptions iused: in.th<! analysis of u'ncontrolled,CEA withdrawal from subcritical co'nditions.

Due.to the high d'iffereritial reac'tivity worth of-the'.shutdown

'CEA groups, the. analysis assumes that,the initial shutdown reiicti"v'ity's such that the react'or,iwill remain subcri'tical. in the event of unexpected" or uncontrolled shutdown group withdrawal..

Other technical specifications ':that reference'the

-Specifications on SHUT-DOWN MARGIN or K..; are: 3/4. 1'.2, BORATION qYSTE(S,, 3/4. I.'3,,

MOVABLE.CONTROL

'SSEMBLIES, 3/4.5.1, REFUIELING OPERATIONS-'E'ORION'CONCENTRATION, 3/4.10.1',

SHUTDOWN MARGIN.AND K.

-:-CEA WORTH TESTS, and 3/4.10.9'SHUTDOWN MARGIN AND K.).-

CEDMS TESTIM.

The 1'imi,tations on moderatoi

'temperature, coefficient (MTC) are provi'ded to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cyc'le'.

The surveillance requirements for measurement of the"MTC du'ring each fuel cycle are adequate t'o con'firm the MTC'alue. since this'oefficient changes slowly due pr.incipally to the reduction in RCS bdroh concentration associ'ated with fuel bu'rnup.

The confirmation'"that the measured

-MTC value i's within its limit prov'ides assurances that the coefficient wil.l be maintained within acceptable values *throughout each fuel cycle; 3 4.1.1.4-MINIMUlhl TEMPERATURE 'FOR CRITICALITY Th'is specifiication ensures t'hat the reactor will not be made critical with the Reactor"Coolant System"cold, 1'eg temperature less than 545-F.

This limita-tion is required "to ensur'e:(1) themodlerator temperature coefficient is within its analyzed temperature ran'ge, '(2)'he -protective instrumentation is, within; its normal operating r'ange;,'(3) ai minimum temperature's p'rovided for Specia'1 Test Except'ion'/4. 10';4;. and '(4) the.reactor vessel-.is above itS,minimum'RT>>,,

temperature.

.PALO VERDE UNIT 1 B'/4 l-.la.

Amendment No. 23-,77,98

3 4.9 REFUE ING OPERATIONS BASES 3

.9.

RON CONC NTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor.vessel.

These limitations are consistent,,with the initial conditions assumed for the boron dilution incident in: the safety analyses.

The boron concentration limit specified in the COLR is based on core reactivity at the beginning, of each cycle (the end of refueling) with all CEAs withdrawn and includes an uncertainty allowance.

This boron concentration limit will ensure a, K,ff of g 0.95 during.the refueling operation.

3 4.9.2'NSTRUMENTATION The OPERABILITY of the startup channel neutron flux monitors ensures that redundant monitoring capability is avail'able to detect changes in the reactivity condition of the core.

3 4.9.3 0 CAY TINE The minimum requirement for reactor. subcriticality pri'or to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

Thi's decay time is consistent with the assumptions used in.

the safety analyses.

3 4.9.4 CONTAINMENT BUILDING PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.

The OPERABILITY. and'losure restrictions are sufficient to, restrict radioactive.material release from a fuel, element rupture based upon the lack of, containment pressurization potential while in, the REFUELING MODE.

3 4.9.5 COM U ICATIONS The requirement for communications capability ensures that refueling station personnel'an be promptly informed of significant changes in the

.facility status or core reactivity condition during CORE ALTERATIONS.

PALO VERDE UNIT I B 3/4 '9-1 Amendment No. 87-.,98

REFUELING OPERATIONS BASES 3/4.9.6

'REFUELING MACHIINE The"-OPERABILITY reqiuiiements for. the 'refue'ling inachine ensure that:

'1) the machine. will be used for in'ovement bf 'fuel assemblies, (2) the-machine has sufficient load capacity to lift a fuel assembly, and (3) the core internals and pressure ves.el are protected from excessive lifting force in the event

,they are i'nadyertently engaged during lift'ing operations';

3/4.9.7 CRANE 'TRAVEL;.,- SPENT FUEL STORAGE'OOL BUILDING The"restriction

-on movement of'oads Iin exces's of the nominal..weightof a fuel assembly, CEA andi associated handling'ool.'ov'er other fuel assemb'lie~s in

the stor'age pool ensures that in the eveht this.load is dropped (1) the, activity release: will be limited tci-that contained in a si'ngle fuel assembly, and'2)'ny possible dlistortion'f fuel. in the storage racks will not. resiult in a critical array,.;This assumpti'on is consistent with. the activity release assumed in the saifety analyses.

3/4. 9. 8 SHUTDOWNI COOLING-AND COBOL/iNT CIRCULATION The requirement, that at 1least oni shutdown'cooling loop'be in operation, and circulating reactor coolant at a flow tate equal to or greater than 3400.gpm (actual) ensures that (1) sufficient cooling capacity is available to remov'e decay heat and maintain the waiter in the reactor pressuire ve!~sel below 1350F-as required during the REFUELIING MODE, (2) suffieienti coolant, circulation is main tained through the, reactor score ito minimize the. effects of' boron dilLitibn incident and. prevent boron,strat'ification, and (3) the AT across the core wi11i be maintained. at 'less thain 75'F during the REIFUELING'MODE.

Th'.iequirald flow'rate'of > 3400 gpm (aictual)'ensures thait at 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months after reactor

'hutdown suf7icient'o'ioling capacity i's avaiilable to remove decay heat 'and

inaintain the water in th'e reactor pressuie vessel. below 135'F's required'uring REFUELING. MODE'; this assumes a shutdown coo] ing heat 'excllianger cooling water.flowrate of 14000.gpma cooling water inlet temperature..if.-<, 1050F'at 27 1/2 hours after reactor shutdown,;aind 'the decay heat curve, of,CESSAR-F Figure 6.2.1-1 and reactor operatioin for two years"at 4000 MWt;

'T'e 37i80i gpm in the specification includes all instrument iincertainties including the 300'F calibration temperature of'he fl'ow transmitters.

Without a shutdown coo1ling train in,op'erati'on'stieam may'be generatedherefore, the 'containment shouldl be sealed'ff to p're.vent escape'f any.

'adioactivity,,and any operatiions that"would cause an incr'i ase in decay'heat should be secured.,

The requirement to have two shutdown cooling loops. OPERABLE when theke lis less than'3 feet of water above the reacto'r pressure

've'ssel flange; en'suf'es'hat a single fai'ilure of the olperat'ing shutdown cooling loop will not res<!it in a complete loss of decay heat removal capak>ility; With the reactor

@esse'1 head, removed and 23,feet, of water above the reactor'ressure, vessel flan'ge, a

large heat sink i. available fior core coplin'g thus in.t,he event, ofai,lure of the operating

. hutdown cooling loop, adelqulte time is provided to initiate emergency procedures to cool tlhe core.

PALO VERDE - UNIT 1 B 3/4 9-2 AMENDMENT NO. 60

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9. 1.9 Core 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:

a.

Shutdown Margin Reactor Trip Breakers Closed for Specification 3.1.1.2 b.

Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.1.3 c.

Boron Dilution Alarms for Specification 3.1.2.7 d.

Movable Control Assemblies - CEA Position for Specification 3. 1.3.1 e.

Regulating CEA Insertion Limits for Specification 3.1.3.6 f.

Part Length CEA Inserti'on Limits for Specification 3. 1.3.7 g.

Linear Heat Rate for Specification 3.2. 1 h.

Azimuthal Power Tilt - T for Specification 3.2.3 i.

DNBR Margin for Specification 3.2.4 j.

Axial Shape Index for Specification 3.2.7,

. k.

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:

"CE Method for Control Element Assembly Ejection Anal'ysis, "CENPD-0190-A, January 1976 (Methodology for Specification 3. 1.3.6, Regulating CEA Insertion Limits).

b.

C.

d.

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

CENPD-266-P-A, April 1983

[Methodology for Specifications 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 (Mod~ 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 (Novenber 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 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 Margin and 3.2.7 Axial Shape Index).

PALO VERDE - UNIT 1 6-20a Amendment No. 69-,7~, 98

41 ADMINISTRATIVE CONTROLS

~EL>>

e.

"Calculative Methods for the CE Large Break LOCA I.'valuation Model for the Analysis of CE and 14 Designed NSSS,"

CENPD-I32, Supplement 3-P-A, June 1985 (Methodology. iFor Specification 3.2.1, Linear Meat Rate).

f; "Calculative iMethods for the CE Smia11

'Break ILOCA Evaluaticin: Model,"

I CENPD-137-P, August 1974 (Methodology for Specification 3.2.1, Linear

-Heat Rate)..

g.

"Calculative Methods for the CE'Smiall B'reaik ILOCA Evaluation Model,",

I, CENPD-I37-P, Supplement 1P, January 1977 (Methodology for Specificatiion 3.2.1, Linear Heat Rate).

h.

-Letter:

0.;D. Parr (NRC) to.F.

M. Stern (CE), dated June '13, 1975 (NRC Staff Review of the CombustiOn Engineering ECCS Evaluation Model).

NRC approval for:

6.9.1.10f.,

i.

Letter.:

K. Kniel (NRC) to A. E.ch'erer (CE),, dated September 27, 1977 (Evaluation os'Topi'ca'1 Report's CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).

NRC appriovhl 'for 6'.9. 1. 10.g.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-miechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear -limits such as shutdown m6rgin, ahd 'transient and anal,ysis

) imits) of the safety analysiis are met.

The CORE OPERATING LIMITS REPORT,:including any mid-cycle revisions or supplements

thereto, shall be provided upon issuance, for each reload cycle, te the NRC. Document Control Deslk with copies'o the Regional Administrator and Resident Inspe'ctor...

PALO VERDE - UNIT 1 6-20b AMENOhiENT NO. ~; 83

eI I'5CO,

,r 0

Cy A,

C p

IIh p

YJ

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++*++

UNITED STATES NUCLEAR'REGULATORYCOMMISSION WASHINGTON, D.C. 20555-0001 R ZONA PUBLIC SERV C

COMPANY ET AL.

.DOCKET NO.

STN 50-529 A

R NUCLEAR GEN RATING STATION UNIT NO.

2 MEN M NT TO FACI ITY OPERATING LICENSE Amendment No.

86 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 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 Cal.ifornia Public Power Authority dated March 31,

1995, complies with the standards and,requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance, (i) that the activities authorized by this amendment can be conducted without endangering the health 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 wil.l not be inimical to the common defense and securi'ty 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:

I I

(2) ical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 86, 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.

The license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION aries 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:

September 1,

1995

l I

1I

ATTACHMENT TG LICENSE'MENDMENT AMENDMENT 'NO.

86 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 identified by amendment number and contain vertical lines indicating the areas of change.

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

REMOVE III*

IV 1-6 3/4 1-1 3/4 1-2 3/4 1-3 3/4 3-'3 3/4 3-7*

3/4 3-8 3/4'-16 3/4,9-1 B 3/4 1-1 B 3/4 1-la B 3/4 9-1 B 3/4 9-2*

6-20a INSERT III*

IV 1-6 4/4 1-1 3/4 1-2 3/4 1-3 3/4 3-3 3/4 3-7*

3/4 3-8 3/4 3-16 3/4 3-16a 3/4 9-1 B 3/4 1-1 B 3/4 1-la B 3/4 9-1 B 3/4 9-2*

6-20a

No changes were made to these pages; reissued to become overleaf pages.

t I

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION 2.1 SAFETY LIMITS

'PAGE 2.1.1, 2.1.1.1 2.1.1.2 2.1.2 REACTOR CORE.....;................

PEAK LINEAR HEAT RATE...........

REACTOR COOLANT SYSTEM PRESSURE.

~ ~

~ ~ ~. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ 0 ~ ~ ~

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0'

~ ~ '

~ '

~

~ ~

~ ~ ~ '

~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

~ ~

~ ~ ~ ~ ~: ~

~

2-1 2 1 2-1 2"1

.2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETPOINTS.................................,.....

2-2 BASES SECTION

2. 1 SAFETY LIMITS:

2.1. 1 REACTOR CORE...............,.......,...........................

2.1.2 REACTOR COOLANT SYSTEM PRESSURE...........................--

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP. SETPOINTS........................-........-

PAGE B 2"1 B 2-2 B 2-2 PALO VERDE - UNIT 2

INDEX

'LIMITING CONDITIONS FOR OPERATION. AND SURVEILLANCE, RE UIREMENTS SECTION PAGNE 3/4 0-1 3/4.1. 1 BORATION CONTROL SHUTDOWN MARGIN REACTOR TRIP BR'EAVERS OPEN.........

3/4 l-l SHUTDOWN MARGIN - REACTOR TRIP.BREAKERS CLOSED.......

3/4 1-2

-MODERATOR TEMPERATURE COEFFICIENT...',..................

3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY,..................

3/4 1-5 3/4-. 1. 2 BORATION SYSTEMS 3/4.1.3 FLOW PATHS SHUTDOWN...............,.....

~.............

FLOW PATHS 'PERATING'............

CHARGiING PUMPS -

SHUTDOWN ~

~

I

~

~ I ~. l.

I ~ ~,'

~

. ~....,

CHARGiING PUMPS OPERATING.... t..I.. l..l...'..'...'........

BORATED WATER SOURCES

-. SHUTDOWN................,......

BORATED WATER SOURCES

-. OPERATING......................

BORON DILUTION ALARMS...............................,

MOVABLE CONTROL ASSEMBLIES 3/4 1-6 3/4 1-7 3/4 'I'-8 3/4:1'-.9 3/4 1-10 3/4 1-12!

3/4 1-13 CEA POSITION..........;....... )....

J.. 3..'.

~

3/4 1-15 POSITION INDICATOR CHANNELS - OPERATING......;......

3/4 1617 POSITION INDICATOR CHANNELS - SHUTDOWN....,..........

3/4 1-18 C EA DROP TIME............,......'..'....'. '...'...'.........

SHUTDOWN CEA INSERTION LIMIT.. 5....

2.. I...'...'.........

REGULATING CEA INSERTION LIMITS.... ~.. ~....,..........

3/4 1-19 3/4 1-20 3/4 1-21 PART LENGTH CEA INSERTION LIMITS......................

3/4 1-23.

PALO.VERDE UNIT 2 Ameridment No4MB, 86

DEFINITIONS REPORTAB E

EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions speci'fied in Sections 50.72 and 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shal,l be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

'No change in part-length. control element assembly position, and.

b.

All full-length control element assemblies (shutdown and regulating) are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

With any full-length CEAs,not capable of being fully inserted, the withdrawn reactivity worth of these full.-length CEAs must be accounted for in the determination of the SHUTDOWN'ARGIN.

SITE BOUNDARY 1.30 The SITE BOUNDARY shall, be that line beyond which the land is neither

owned, nor leased, nor otherwise controlled by the licensee.

SOFTWARE 1.31 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated

data, documentation, and procedures..

SOURCE CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor.

is exposed to a source of increased radioactivity.

STAGGER D T ST BASIS 1.33 A STAGGERED TEST BASIS shall consist of:

a.

A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified'est interval into n equal subintervals, and b.

The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.

THERMAL POWER 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

PALO VERDE UNIT 2 1-6 Amendment No. k3-;48, 86

0 h

I

~ '

~

'EACTIVITY CONTROL SYSTEMS 3 4.1 REACTIVITY CONTROL SYSTEMS 3 4. 1. 1 BORATION CONTROL SHU DOWN MARG N R ACTOR TRIP BREAKERS OPEN**

LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0X delta k/k.

5 TT: 'IIDDE,A 55 Itk 5 <<I 5

k ACTION:

With the SHUTDOWN MARGIN less than 1.0X delta k/k, 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.

SURVEILLANCE RE UIREMENTS

4. l. 1. 1. 1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0X del'ta k/k 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

.fol.lowing 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 shal51 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. 1. 1. 1, above.

The predicted reac-tivity 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'ithin 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. 48, B6

S

~P<<L Y!I I SHUTDOWN MARGIN REACTOR 'll'RIP FiREAKERS CLOSED*'~

LIMITING CONDITION FOR OPEF(ATION 3.1.1.2 a ~

b.

c ~

The SHIIJTDOWN MARGIN shall be greater thain or equal to that specifiedI in the:CORE OPERATING LIMITS REPORT, and For T,<< le. s than or equal to 500'F, K, shall be less than'.'99'.

Reactor. criticality shal'1 not be achieved with shutdown group CEA movemeint.

~PI

'PP IIIPPPI., ",,

PP>>IP:P

. 'PP PP'*

closed**.

~CT ON:

a ~

b.

With t'e SHUTDOWN MARGIN less than that specified in the CORE OPERATING LIMITS REPORT, immediately initiate and continue, b'oration at greater than or equal to 26 gpm to the reaictor coolant sy'tem of a solution containing greater than or equal to 40130 ppm boron or equiva'lent unPtiT the required SHUTDOWN MARGIhl is restoredand With T,,<<, less than or equal to 500'F and I(., greater than or <qual, to 0.9!), immediately vary CEA positions and/air initiate and continue boration at greater than or e'qual to 26 gpmP to, the reactor c'oolant system of.a solution con'taining greater than or equal to 4000 ppm boron or equivalent unti'1 the required K, is, restored.

SURVEILLANCE RE UIREMENTS

4. 1. 1.2. 1

,With the reactor trip. breakers Icldsed*+, the SHUTDOWN MARGIN shall i

be determined to be gr'eater than or equal Ito that speci,'fied in the CORIE OPERATING LIMITS REPORT:

a.

Within 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 ati least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , thereafter whi'le the CEA(s), is inoperable.

See Special Test Exceptions. 3.10,.1 and 3.10.9I.

- The CEA drive system capaibl,ie of CEA,withdrawa'l.

'ALO VERDE UNI I' 3/4 1-.2 Amendment 'No. R-,+5P6

ACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREHENTS Continued b.

When in MODE 1 or MODE 2 with k

~ greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within the Transient Insertion Limits of Specification 3. 1.3.6.

If CEA group withdrawal is not within the Transient Insertion Limits of Specification 3. 1.3.6, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months verify that SHUTDOWN HARGIN is greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

c.

When in MODE 2 with k,> less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor critYcality by verifying that predicted critical CEA position is within the limits of Specification 3.1.3.6.

d.

Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e.

When in MODE 3, 4, or 5, 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. 1.2.2 When in MODE 3, 4, or 5, with the reactor trip breakers closed**,

and T,i less than or equal to 500 F, K, shall be determined to be less than 0.99 aE feast once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at least the fol.lowing 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. 1.2.3 When in MODES 3, 4, or.'

with the reactor trip breakers closed**,

verify that criticality cannot be achieved with shutdown group CEA withdrawal 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. 1.2.4 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. 1.1.2. I.e or 4. 1. 1.2.2.

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.

- The CEA drive system capable of CEA withdrawal.

PALO VERDE UNIT 2 3/4 1-3 Amendment No. 48,86

MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3'.1.-1.3 The moderator temperature cdefficient (HTC) shall be. within the area of Acceptable Ogieration speciified in the CORE OPERATING LIHITS4REPORT,,

The maximum upper limit sha'll be less than or equa'( to + 0.5 x 10 AK/K/ F fop a powet level of CIX RATED THERMAL PCtWER with a linear.,ramp to OAK/K/'F at 100X

'RATED THERMAL PCIWER, APPLICABILITY:

MODES 1 and 2"¹.

ACTION:

With the moderator temperature coefficierit outside the area of Acceptable, Operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE(~UIREtlENTS 4.1. 1.3.1 The HTC shall be determined to be withiri its limits by confirmator'y measurements.

HTC measured Ivalues shall be extrapo'late'd and/or compensated to permit direct compariison with the above limits.

4. 1. 1. 3. 2 The HTC shal 1 be determined at the foll'owing frequencies and THERMAL POWER conditions duriing each fuel cycle:

a.

Prior to.initial operation'bove SX of iRATED'HERMAL POWER,'afte'r each fuel loading.

b.

At any THERMAL POWER; within 7 EFPD after reaching a core a'verage exposure of 40 EFPD burnup into the current,cycle.

C.

At. any THERMAL POWER, w'ithin 7 EFPD after reaching a core average exposure equivalent to two"thirdsjof the expected current cycle end-of-cyc1e core average burnup.

With K.ff greater than or equal.to 1.().

eff

¹See Special Test Exception,'3.10.2.

PALO VERDE - UNIT 2 3/4 1-4 AMENDMENT NO.

FUNCT ONAL UN I.

TRIP GENERATION A.

Process TABLE 3.3-1 R ACTOR PROT CTIVE NSTRUHENTATION HINIHUH TOTAL NO.

CHANNELS CHANNELS OF CHANNELS TO TRIP OPERABLE APPLICABLE

~OO S

OCTION 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 8.

Local Power Oensity - High 9.

DNBR - Low B.

Excore Neutron Flux 4

4/SG 4/SG 4/SG 4

4/SG 4

2 3

2 (b) 3 2/SG 3/SG 2/SG 3/SG 2/SG 3/SG 2

3 2/SG 3/SG 2 (c)(d) 3 2 (c)(d) 3 1,

2 1,

2 3*

4*

1, 2

1; 2

1, 2

2¹', 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹', 3¹ 2¹, 3¹ 2¹, 3¹ 1;

Variable Overpower Trip 2.

Logarithmic Power Level - High a.

Startup and Operating b.

Shutdown C.

Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators 2 (a)(d) 3 2 (e) 2 (c)(d) 3 1,

2 2¹, 3¹ 1,

2 2¹, 3¹ 3*

4*

5%

9 3,4,5 4

1, 2

6, 7

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

2¹, 3¹, 7, 10

41

~

Q

~

I~

I

TABLE 3. 3-1'Continued)

REACTOR PROTECTIVE INSTRUMENTATION ACTION'STATEMENTS ACTION 4 ACTION 5 3.

Steam Generator Pressure -

Steam Generator Pressure

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

Steam Generator Level 2-Low (ESF) 4.

Steam.Generator Level' L'ow Steam Generator Level - Low (RPS)

(Wide Range)

Steam Generator Level 1-Low (ESF)

Steam Generator Level 2"Low (ESF) 5.

Core Protection Calculator Local Power Density - High (RPS)

DNBR' Low (RPS)

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

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

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requi,rement,,

suspend all operations involving, positive reactivity changes.

With the number of channels OPERABLE one less than -required, by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise,'e

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

however, the trip breaker associated'ith the inoperable channel may 'be closed for.up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for surveillance testing per Specification 4.3.1.1.

ACTION 6

a.

With one CEAC inoperable, operation.may continue for up to 7 days provided that the requirements of Specification

4. 1.3. l. 1 are met.

After 7 days,,operation may continue provided that the conditions of Action Item 6.b are met.

b.

With both CEACs inoperable, operation may continue provided that:

1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the DNBR margin required by Specifica-tion 3. 2. 4. b (COLSS in service) or 3'.2.4. d (COLSS out of service) is satisfied and the Reactor Power Cutback System is disabled, and PALO VERDE - UNIT, 2 3/4 3"7 AMENDMENT NO. 36

ACTION 7 ACTION 8 REACI'OR PROTECTIVE INS'll'RU, ENTA'TION fiCTION STATEMENTS 2.

Within 4 Ihours:

a)

All full-'length and part'-length CEA groups must be withdrawn w'ithin the limits of Specif'ic4ti0ns 3.,1.,'3.5,

3. 1.3,.6b, and 3.1.3.7b, except during su} veillance testing pursuant to the requirements of Specification 4. 1.3.1.2.

Specification 3 1.,3.6b allows CEA group 5 insertion to no further tlhan 127.5 inches withdrawn.

b)

The "RSPT/CEAC Inoperable" addressable con. tant in the CPCs is set to indicate. that both CEACs are inoperable.

c).

The Control Element Drive Mechanism Control System (CEDHCS) is placed iin aind subsequ'ently'aintained in 'the '"Standby" mode except during CI=A motion pi'ermitted by.Specifications 3.1..3.5, 3,.1.3.6b and 3.1.3.7b when the.

CIEDHICS may be operated in either the "Hanual Group" orl "Manual Individual" mode.

3.

CIEA position survei'llance must meet, tlhe requir erinents of Specifications

4. 1..3. 1. 1, 4. 1.3.5, 4.1.3.6, and 4.1.3.7 except du}ing surveill,ance testing pursuant to Specification 4.1.3.1,.2.

With tlhre,e or more auto restarts; exclueiing pieriodic auto rest;arts,(Code 30 and Code 33), of one non-bypassed cal',cu)at'or',

.during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL.IFUNCTIONAL TEST within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the number of iOPERABLE channels one less than the Minimum Clhanne'Is OPl;RABLE requirementrestore ann inoper'able chanr)el, to OPEIVBLE status iwithiii 48 iiours or opien an affected reactor t'ripe b>reaker within the next hour.

ACTION 9 With the number of IOPERABLE channels one, less than the Minimum Clhannels OPERABLE requirementrestore t,he inoperable char)nel to OPEIVBLE status iwithin 48 t>ours or opien the reactor trip breakers within the next hour,.

ACTION 10 In HODIES 3, 4, or 5, the. Core Protection Calculator channels are not required to be iOPERABLE when the Logarithmic Power Level,,

High trip is OPERABLE with the trip setpoint lowered to g10 'X of RATIED THI=RMAL Power.

PALO VERDE UNIT 2 3/4 3-8 Amendment; No. 36,86

TABLE 4.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREHENTS FUNCTIONAL UNIT D.

Supplementary Protection Syste~

Pressurizer Pressure

- High CHANNEL CHANNEL CHFCK CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE ED, 1,

2 I ~

II.

RPS LOGIC A.

Matrix Logic B.

Initiation Logic I II.

RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manual Trip N.A.

N.A.

M, R(10) 3*

4*

5*

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

3*

4*

5+

C)

~RACTOR PROjT:CT!UE IIISTRUNENTA 10 UR EILLANCE RE VIRENENTS, (2)

TABLE NOTATIONS With reactor trip breakers in, the closed posi,tion.and'he CEA drive, system capable of CEA withdrawal,,and fuel in the reactor vessel.

Each 'STARTUP or when required witth ~the reactor trip breakers closed and the CEA drive s'stem capable of,rod withdrawal, if not performed in the previous 7 days.

Heat balance only (CHANNI=L FUNCTIONAL TEST not included):

a.

Between 15X and 80X of'ATED lHERMAL POWER,,

compare the linear power level, the CPC delta T power and the,CPC nuclear

power, signals to the calorimetric calculation.

If any signal is within -0.5X to 10X of the calorimetric then do not calibrate except as reguired during initial power

a. cension after refueling.

If any signal is less than the calorimetric calculation by more than 0.5X,, then adjust th!e affected signal(s) to agree with the calorimetric calcu'lat,ion.

If any signal is greater than the calorimetric cal!culation by more than 10X then adjust the afFected signal(s) to agree, with the calorimetric calculation w!ithin 8X to 10X.

b.

At or above 80X of RATED THEfNtAL.POWER; compare the linda'bweir level, the CPC delta T power and the CPC nuclear power signals,to the calorimetric calculation.

If any signal differs from the calo'rimetric calculation by a6 absolute difference, of more than 2X, then adjust the affected signal(s) to agree with the calorimetric calculation.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding 1to the next major test power plateau.

(3)

Above 15X of RATED THERMAL POWER, verify that the linear power sub-channel gains of the excore detectors are consistent with the.values used to establish the shape an!nealing matrix !elements in the Core Protection Calculators.

(4)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5)

After each fue'I loading acid prior to exceeding 70X of RATED THERMAL

POWER, th!e incor e detectors shall be ised to determine the shape annealing matr'ix elements and the Core Protection Calculators shall use these elements.

PALO VERDE = UNIlI 2 3/4 3-16 Amendment No. 38-,IW, 86

TABLE 4.3-1 Continued ACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS TABLE NOTATIONS (6)

This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel as cl'ose to the sensors as practicable to verify OPERABILITY including alarm and/or trip functions.

(7)

(8)

Above 70X of RATED THERMAL POWER, verify that the total steady-state RCS flow rate, as indicated by each CPC is less than or equal to the actual RCS total.flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by,calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow, is less than or equal to the actual flow rate.

The flow measurement uncertainty may be included in the BERRl term in the CPC and is equal to or greater than, 4X.

Above 70X of RATED THERMAL POWER, verify that the total, steady-state RCS flow rate as.indicated by each CPC is less. than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation and the.ultrasonic flow meter adjusted pump curves or calorimetric calculations.

(9)

The quarterly CHANNEL FUNCTIONAL TEST shall include verification that the correct (current) values of addressable constants are installed.

in each OPERABLE CPC.

(10)

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

PALO VERDE UNIT 2 3/4 3-16a Amendment No.86

41

'1 1'

3 4.9 REFUELING OPERATIONS 3 4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9. 1 With the reactor vessel head closure bolts less than fully tensioned or.

with the head

removed, the boron. concentration of all fil:led portions of the Reactor Coolant System. and the refueling canal shall:be maintained uniform and within the limit specified in the Core Operating Limits,Report (COLR).

~AI TT:

EOOE I'.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 26 gpm of: a solution containing

> '4000 ppm boron or its equivalent until the boron concentration is within limits.

SURVEILLANCE RE UIREMENTS 4.9. 1. 1 The boron concentration shall be determined to be within the limit specified in the COLR prior to:

a.

Removing or unbolting the reactor vessel head',

and b.

Withdrawal of any full-length CEA in excess'f 3 feet from its fully inserted position within the reactor pressure vessel'.

4'.9'. 1.2 The boron concentration of the Reactor Coolant System a.d the refueling canal shall be determined by chemical analysis at,least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in NODE 6.whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned:

or with the head removed.

-PALO VERDE UNIT 2 3/4 9-1 Amendment No. 89,B6

REFUELING OPERATIONS 3/4. 9. 2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3;9.2 As a minimum, two star tup channel neutron f'lux monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and, one with audible indication in the containment and control room.

APPLICABILITY:

MODE B.

ACTION:

a.

b.

With one of the above required monitors inoperable or not operating,'mmediately suspend all operations involving CORE ALTERATIONS or positive reactivity ctianges.

With both of'he above.required monitors inoperable or not operating, determine the boron concentration, of the Reactor Coolant System at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .,

SURVEILLANCE RE UIREMENTS 4.9.2 Each s artup channel neutron flux moni'tor sha'll,be,demonstratedi OPERJBL'E by 'performance of:,

a.

A CHANNEL CHECK. at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

A CHANNEL FUNCTIONAL TEST within 8 hourg pliar to the initial start of CORE ALTERATIONS, and c.

A CHANNEL FUNCTIONAL TESI at leastlonCe per 7, days.

PALO VERDE " UNIT 2 3/4 9"2

3 4.1 REACTIVITY CONTROL SYSTEMS BASES 3 4. 1. 1 ORATION CONTROL 3 4. 1. 1. 1 and 3 4. 1. 1:2 SHUTDOWN MARGIN and K

The function of SHUTDOWN MARGIN is to ensure that the reactor remains subcritical following a design basis accident or anticipated operational occurrence.

The function of K., is to maintain sufficient subcriticality to preclude inadvertent criticality following ejection of a single control element assembly (CEA).

During operation in MODES 1 and 2, with k << greater than or equal to 1.0, the transient insertion limits of Specificatson 3.1.3.6.ensure that sufficient 'SHUTDOWN MARGIN is available.

SHUTDOWN MARGIN is the amount by which the core is subcritical, or would be subcritical immediately following a reactor trip, considering a single malfunction resulting in the highest worth CEA failing to insert.

K., is a

measure of the core's reactivity, considering a single malfunction resulting in the highest worth inserted CEA being ejected.

SHUTDOWN MARGIN requirements vary throughout the core life as a function of fuel depletion and reactor coolant system (RCS) cold leg temperature (T,).

The most restrictive condition occurs at EOL, with T,~ at no-load operating temperature, and is associated with a postulated steam line break accident and the resulting uncontrolled RCS cooldown.

In the analysis of this

accident, the specified SHUTDOWN MARGIN is required to control-the reactivity transient and ensure, that the fuel performance and offsite dose criteria are satisfied.

As (initial) T ~ decreases, the potential RCS cooldown and the resulting reactivity transient are less severe and, therefore, the required SHUTDOWN MARGIN also decreases.

Below T,~ of about 350 F, the inadvertent deboration event becomes limiting with respect to the SHUTDOWN MARGIN require-ments.

Below 350'F, the speci'fied SHUTDOWN MARGIN ensures that sufficient time for operator actions exists between the initial indication of the deboration and the total loss of shutdown margin.

Accordingly, with the reactor trip breakers closed and the CEA drive system capable of CEA withdrawal, the SHUTDOWN 'MARGIN requirements are based upon these limiting conditions.

Additional events considered in establishing requirements on SHUTDOWN MARGIN that are not limiting with respect to the Specification limits are single CEA withdrawal and startup of an inactive reactor coolant pump.

K., requirements vary with the amount of positive reactivity that would be introduced assuming the CEA with the highest inserted worth ejects from the core.

In the analysis of the CEA ejection event, the K., requirement ensures that the radially averaged enthalpy acceptance cr.iterion is satisfied, considering power redistribution effects.

Above T,~ of 500,F, Doppler reactivity feedback is sufficient to preclude the need for a specific K.,

requirement.

With all CEAs fully inserted, K., and SHUTDOWN MARGIN requirements are equivalent in terms of minimum acceptable core boron concentration.

PALO VERDE UNIT 2 B 3/4 1-1 Amendment No. 4S, 86

BASES Jill

' "',-I-'"

'1

.The requireimeht prohiliitinc~ criticality:due to shutdown group CEA movement is associated with the assumpti(ins used in, the analy'sis of uncontrolled CEA withdrawalfrom subcritical.conditions.

Due.'to -the, hicjh differential reactivity,worth of, the.shutdown,CEA'. groups, the. analysis,assume's that the ini.ti'al shutdown re'activity is such, that, the reactor, will, rem'ain subcritical i'

<<theT event of-unexpected or. uncontrolled shutdown. group withdrawal.

Other..technical'peci Ficat;ions. that reference the,.Specifi'cations on SHUTDOWN MARGIN or K are:

3/4. 1.2, BORATION SYSTEMS;.3/4. 1;3,,

MOVABLE CONTROL ASSEMBLIES, 5)4.9.1, REI:UELING DPERATIQNS-BOROII CONCEN'TRATION, 3'/4.- 10. 1, SHUTDOWN-MARGIN,.AND.;,Kc, -CEA WORTH TESTS, and 3/4. 10.9,.

SHUTDOWN.

MARGIN AND K-: CEDMS IIES'TING;

<<~I The limit'ations on moderator teiiiperature coe1F'ficii!nt (MTC). ar'e provided, to, ensure that -the assumptions used. in the. accident,and,the transient

analysis, remain valid through.each fuel cycle.

'The surv'eillance, requ'irements for measurement, of the MTiC dluring each 'fiiel.cycle are -adequate to.confirm the MTC value, since.this cciefficieii't changes.slowly due principally to..the, reduction in

-'RCS boron concentration, associated with fuie1 burnilp..The,.confirma'tion that the measured 'MTC,.value is with'in its limit prqvides assurances

-that. the coefficient

.will be.m'aint'ained within a'cceptable va'lues, throughout each.-fuel cycl,eI.

Thi's specifi'cation. ensures that the reactor, will,,riot be m~ie critical with the Reactor Coolant,:..System cold leig.temperature,.less thian 545'F,.

This l.imitation is required to, ensure (1) the moderator temperature -coeffi'cient;is within its analyzed-temper'ature range,,(2)-;the.protective,iiiistrumentati'on is within;its norma'1.operating

range, (3) -a min',imum temPperature,.is proyid'ed'fo'r Special: Test:Except,ioh 3/4 10.4, and (4) the reactor,.vessel is above i.ts minimum RT>> temperature.

PALO VERDE

, UNIT 2 B 3/4 1-lai Amendment No~ 48-68,86,

3 4.9-REFUELING OPERATIONS BASES 3 4.9. 1 BORD CONCENTRATION The limitations on reactivity conditions during. REFUELING ensure that:

(1) the reactor. will remain subcritical dur'ing CORE ALTERATIONS,.and: (2) a uniform boron concentration is maintained for reactivity control, in the water volume having direct access to the reactor vessel.

These l,imitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.

The, boron concentration, limit specified in the COLR is

'based on core reactivity at the.beginning of each. cycle (the end of refueling) with all: CEAs withdrawn and includes an uncertainty allowance.

This. boron concentration l,imit will. ensure a K,<< of, ~0.95 during the, refueling. operation.

3 4.9.2 INSTRUMENTATION The OPERABILITY of the startup channel neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3 4.9.3 DECAY TIME, The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assembl.ies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products.

This decay: time,is consistent with the assumptions used in the safety analyses.

3 4.9.4 CONTA NMENT BUILDING PENETRATIONS The requirements on containment, penetration closure and-OPERABILITY ensure that a release of radioactive material'ithin containment will: be restricted

.from leakage to the environment.

The OPERABILITY and closure restrictions are sufficient, to. restrict radioactive materi'al release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.

3 4.9. 5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

PALO VERDE UNIT 2 B 3/4 9-1 Amendment No.

B6

REFUELING OPERATIONS Il

~

g BASES 3/4. 9. 6 REFUELING MACHINE The OPERABIILITY irequirements for tlhe refuel,iiig-machine ensure that:;

(1) the machine. will be used fo>

movement. of fuel assemblies, (2).the machine has, sufficient load.capacity to l.ift a fuel a'ssembly, and (3).the core internals and pressure;.vessel aI;e prdtected from excessive 'lift'ing fot"'ce 'in the Ieventi, they 'are inadvertenitly engaged.',durinq lifting operations.

3/4.9.7'RANE TIRAVEL ": SPENT -FUEL STORAGE POOL BUILDING The restriction on moveme'nt of. loads i'n excess of, the nominal w'eight of a fuel"assembly, CIEA and associated

'handl.ing. tool over other fue'l. assemblies.;in; the storage -pool-ensures that in the event. this lead is dropped'(1) tHe

-,activity release will be limited-to that contained in a sinqle fuel assembly, and (2) any possible distortion of fuel iri the storage, racks will not redullt

'n a critical; array.

This assumption is consistent. with the activity,re'lease,

-assumed; in the.safety analyses.

3/4. 9. 8'HUTDOWN 'COOLING AND COOLANT 'CIRC'ULATION

'The requirement that at least one shutdowni cooling loop. be in operation,,

and circulating reactor coolant at a flow irate equal,to or greater thaln 3400 gpm (actual) ensures that (1) sufficient coo1ling capacity is available, t6 remove; decay heaI> and maintaiin the water in the reactor pressure vessel below 1354F, as.requireIf duriing ttIe REFUEILING MODE, (2) sufficieirit co'olant circulation is maintained through the reaictor core to minimize the effects of,a-boron dilution,incident andprevent boron stratification, and (3) the bT across 'the, core. will'be maintained at less than 75~F dukinlg the'EFUELING'MODE;

'The required flowrate of > 3400i'pm (actu'al) ensures that ait 288 'hours'fter reactor shutdown sufficient cooling capacity is available to remo've decay heat and-maintain the water in the reactor pressure vessel below.135~F as required during REFUELING MODE; this assume.

a shutdown cooling heat e'xchanger..cooling water flowrate of 1400!0 gpmi, a coo'ling water inlet, temperature of < 1050F.at 27 1/2 hours -after'eaictivr shutdown; and the"decay h! at curve of CESSAR"F Figure 6. 2.1-1 and reactor operatibn for two years at 4000.MWt.

The 3780 gpm

-in the specification includes all ',instrwme'nt 'uncertainties including the 3004F calibration 'temperature bf the flow transmitters.

Without a shutdown cooling-train in operation steam may be genl.rated;

'herefore, the, containment should be sealed cIff to prevent escape of any,',

radioactivity, and any ojverations that would cause an.increase in decay heat

-should be secured.

The requirement t'o have two,sliutdow'n cooling loops OPERABLE when,there,'s less than 23 feet of Qater above the re<<ictor priessure vessel

flange, ensures that a s'ingle-failure of thie operating 0hutd!Iiwn cdoling loop will not',resul't, in a complete loss of decay heat removal capability.

With the reactor

.vess'el'head removed and 23 feet of-water above'th'e reactor pressure vessel

flange, a

large, heat sink is available for core cool'ing, thus in the event of a failure of the operating shutclowIn coo'ling 'loop, adequate time is provided to initiate emergency procedures to icool 'the core.

'PALO'VERDE - UNI1' B 3/4 9-2 AMENDMENT NO. 47'

'ADMINISTRATIVE CONTROLS CO 0

RATING HITS REPORT 6.9.1.9 Core operating limits shall be establ,ished and documented in the CORE OPERATING L'IMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

a ~

b.

C.

d.

e.f.

g.

h.

k.

Shutdown Margin 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 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. 1.1.2, Shutdown Margin Reactor Trip Breakers Closed;

3. 1. 1.3, Moderator Temperature Coeffici'ent 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 (Novenber 1981),

Supplements No.

1 (March 1983),

No.

2 (September 1983),

No.

3 (December 1987)

(Methodology for.Specifications

3. 1. 1.2, Shutdown Hargin 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, Hay 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.2 6-20a Amendment No. 5S-';68,86

S

'"i "' M!-

t '"~l"')

~i e.

Calcu'1ative Nethod. for the CE Large 'Break LOCA.Evaluation Model for the Analysis of CE and M Dedig'ned NESS,"'CENPD-132, Supplement 3-P;-A, June 1985 (MetIhodology for Specification 3.2.1, ILinear Heat R'ate).'.

Calcu'lative Methods for the CE Small Break LOCA Evaluation Model,i'ENPD-137-P A>gust 1974 (Methodology 'for Specification 3.2.1, Linear Heat Rate).

9-h.

"Calculative Nethods for the CE Small Break LOCA Evaluation Hodel,"

CENPD-137-P, Supp'lement 1P, January 1977 (Methodology for Specificaticin 3.2.1, Linear Heat Rate).

Letter::

O. D. Parr (NRC) to F'. H. Stern '(CE), dated June 13, 1975 (NRC Staff Review of'he Combu.stion Engineering ECCS Evaluation Medel).

INRC'approval for:

6.9.1.10f.

i.

Le'tter:

K. Kniel (NRC) to A. E. Scherier (Cl=), dated September 27, I

1977 (Eva'1uation of Topical Reports CEINPDI-133, Sup"lement 3-P'nd CENPD-137,,

Supplement 1'-P).

NRC approval for 6.9.1.10.g I

The core operating limits shall be determined so that all app'licable ~limits (e.g., fuel thermal-meehan'ical limits, core thermal-hydraulic limits, ECCS limits, nuclea>r lim'its such as shutdown margin, and transient and ana!lysis, limits) of tSe safety analy'is are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements

thereto, shall be provided upon issuance, for each reload cycle, to the NRC Oocciment Control Desk with copieC to the Regi{)nal Administrator and Resident Inspector.

PALO VERDE UNIT 2 6-20'MFNDMENT NO. -95, 70

~PS AE00 Wp0 ty p%

g) 00

~O

+w*e+

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D;C. 2055$ 4001 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. 69 License No.

NPF-74 The Nuclear Regulatory Commission (the Commission) has found that:

A.

The appl.i cati on for amendment by the Arizona Public Service Company (APS or the licensee) on behal.f of itself and.the Salt River Project Agricul tural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Depar tment of Water and

Power, and Southern California Publ ic Power Authority dated March 3 1, 1 995, compl ies with the standards and requirements of the 'Atomi'c Energy Act of 1 954, as amended (the Act) and the Commi ss i on '

regulations set forth in 1 0 CFR Chapter I; B.

The facility will operate in conformity with the application, the provi s ions 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 Commi ss ion '

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 al,l applicable requirements have been satisfied.

2.

Accordingly, the -license is amended'y 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:

~

f 11

~ '

(2)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

69, 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.

The license amendment is effective as of the date of.issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Charles 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:

September 1,

1995

~

)

y 4l E

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

69 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.

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

REMO E

III*

IV 1-6 3/4 1-1 3/4 1-2 3/4 1-3 3/4 3-3 3/4 3-7*

3/4 3-8 3/4 3-16 3/4 9-1 B 3/4 1-1 B 3/4 1-la B 3/4 9-1 B 3/4 9-2*

6-20a INSERT III*

IV 1-6 4'/4 1-1 3/4 1-2 3/4 1-3 3/4 3-3 3/4 3-7*

3/4 3-8 3/4 3-16 3/4 3-16a 3/4 9-1 B 3'/4 1-1 B 3/4 1-la B 3/4 9-1 B 3/4 9-2*

6-20a

No changes were made to these pages; reissued to become overleaf.pages.

r

~

0 h ~

~,

~ ~

INDEX SAFETY LIMITS AND 'LIMITING SAFETY SYSTEM SETTINGS SECTION 2.1 SAFETY LIMITS PAGE 2.'1. I 2.1.1.1 2.1.1.2 2.1.2 REACTOR CORE....................

.PEAK LINEAR HEAT RATE...........

REACTOR COOLANT SYSTEM PRESSURE.

~ ~ ~ ~

2 1

~ ~ ~ ~

2 1

~ ~ o ~

2 1

~ ~ ~ o 2 1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETPOINTS.........,.............-...........".

2-2, t

BASES SECTION 2.1 SAFETY LIMITS

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE.............................

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP, SETPOINTS.....................................

PAGE B 2-1 B 2-2 B '2-2

'PALO VEROE - UNIT 3 IIX

g

~

INDF.X LIMITING CONDITIONS FOR OPERA'TION AND SURVEIlLLA CE RE UIRf:MENTS

~SC 10M

! 'l

!!'/4.

1. 1 BORATION CONTROL

~

PAGE 3/4 O-l, SHUTDOWN MARGIN - REACTOR TRIP BREAKERS'PEN..........

SHUTDOWN MARGIN REACTOR TRIP. BREAKERS CLOSED........

3/4 1-1 3/4 1-2 MODERATOR TEMPERATURE COEFFICII=NT...................

3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY............

'3/4 1'-5

'/4.

1. 2 BORATION SYSTEMS FLOW PATHS SHUTDOWN.........

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

FLOW PATHS OPERATIING..........

CHARGING PUMf'S SHUTDOWN.................;...........

,3/4 1;6

,3/4 1;7 3/4 1;8 CHARGING PUMPS - OPERATING..i.. L. i..'.

........ 3/4 1;9 BORATED WATER SOURCES - SHUTDOWN.i........

3/4 1-10 BOILED WATER SOURCES - OPERATING...i..................

3/4 lt-l2 BORON DILUTION ALARMS.......',...'..'......................

3/4,1.3 MOVABLE CONTROL ASSIEMBLII:S CEA POSITION........,......,............,...............

POSITIOhl INDICATOR CHANNELS J-OPERATING..'..

POSITIOhl INDICAI'OR CHANNELS i-SHUTDOWNS..'..

CEA DROP TIME 3/4 l-l3 3/4 1-15 3/4 1+17 3/4 lr18 3/4 1-19 SHUTDOWN CIEA INSERTION LIMIT'...'..'.................

3/4 1-2O REGULATING CEA INSERTION LIMITS.....,.................

3/4 1~21 PART LEN'GTH CIEA INSERTION LIMITS............,..

~

3/4i 1-23 i

PALO VERDE UNIT 3 IV Amendment No; R3-,4&,69

DEFINITIONS REPORTABLE EVENT 1.28 A 'REPORTABLE EVENT shall be any of those conditions specified in Sections 50.72 and 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.29 SHUTDOWN 'MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a.

.No change in part-length. control element assembly position, and b.

All'. full-length control element assemblies (shutdown and regulating) are fully i'nserted except for the single assembly. of highest reactivity worth which is assumed to be fully withdrawn.

With any full-length 'CEAs not capable of being fully.inserted, the withdrawn reactivity worth of these full-length CEAs must be accounted for in the determination of the SHUTDOWN. MARGIN'.

SIT BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither

owned, nor leased, nor otherwise controlled by the licensee.

~SOFTMAR 1.31 The digital computer SOFTWARE for the reactor protection system shall be the.program codes including their associated

data, documentation, and procedures.

SOURC CHECK 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

STAGGER D T ST BASIS 1.33 A STAGGERED TEST BASIS shall. consist of:

a.

A test schedule for n systems,.

subsystems,

trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and b.

The testing of one system,,

subsystem, train, or other designated component at the beginning of each subinterval.

~A 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor.coolant.

PALO VERDE - UNIT 3 1-6 Amendment No. 8-,34, 69

ACT V TY C

TROL SYSTEMS 3 4.1 R ACTIVITY CONTROL SYSTEMS 3 4..1 ORATIO CONTRO SHUT OW MARG N REACTOR TRIP BREAKERS OPEN**

LIMITING CON ITION FOR OPERATION

3. 1. 1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0X delta k/k.

'NDE;4 d

~ih h

<<ipb k

~CTI ON:

With the SHUTDOWN MARGIN less than 1.0X, delta k/k, 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.

SURVEILLANCE RE UIREMENTS

4. 1. 1. 1. 1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0X delta k/k 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.

5.

6.

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

Xenon concentrat'.on, 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 delta k/k at least once per 31 Effective Full Power.

Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4.1.1. 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 3 3/4 1-1 Amendment No. 8, 69

fe SHUTDOWN MARGIN - REACTOR TRIP BREAKERS CLOSED*"

LIMITING 3.1.1.2

'a ~

C.

CONDITION FOR OPERATION The SHUTDiOWNI MARGIN shall be greate'r thain or equal to that specified in the CORE OPERATING LIHITS REPORT,,

and For T<< 'less than or equal to 500 F, K, "hall be less than 0.99 Reactor criticality shall'ot be achieved with shutdown group

CEA, movement.

82LIIELLII:

E, ', 'E',

E '

l R

R P

C closed**.

~CTI ON:

a.

lith the SHUTDiDWNI MARGIN less than that specified in the CORIE OPERATING LIMITS REPORT, immediatel'y initiate and continue boration at greater than or equal-to 26 gitrm Itol th'e reactor coolant system of a solution containing greater than or equal to 40()0 ppm boron or equiva1lent until the required SHUTDOWN MARGIN i

restored, and b.

With T,<< less than or eclual to 500 F and K', greater than or equal to 0.9V, immediately vary CEA'opitiolss,and/or jnitiate and continue boration at greater than or equa'l to 26 gpm to the reactor cool,ant system of a so'tution containing greater than or equal to 4000 ppm boron or equivalent until the required K, is restored.

~LL EE

4. 1. 1.2. 1 Pith the reactor. trip breakers closed**, the SHUTDOWN MARGIN sha'ill be determined to be greater than or equal to that specified in the CORI=

OPERATING LIHITS REPORT:

a ~

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detectiion 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 IExceptions 3.10.1 and 3.10.9.

- The CEA drive system capable of CEA withdrawal.,

PALO VERDE UNIl 3 Amendment Nlo. 2-,48,69

ACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued b.

When in MODE 1 or MODE 2 with k

~ greater than or equal to 1.0, at least once per 12 'hours by veri/ying that CEA,group withdrawal is within the Transient Insertion Limits of Specification 3. 1.3.6.

If CEA group withdrawal is not within the Transient Insertion Limits of Specifi'cation 3. 1.3.6,, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months verify that SHUTDOWN MARGIN is greater than or equal to that specified in the CORE OPERATING LIMITS REPORT.

c.

When in MODE 2 with k,> less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criC>cality by verifying that predicted'ritical CEA position is within the limits of Specification 3.1.3.6.

d.

Prior to initial operation above 5X.RATED THERMAL POWER after each fuel loading, by consideration of the factors of e.

below', with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e.

When in MODE 3, 4, or 5, 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 concentrati'on.

4. 1. 1.2.2 When in MODE 3, 4, or 5, with the reactor trip breakers closed**,

and T<< less than or equal to 500'F, K<.

shall be determined to be less than 0.99 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by cons>Jeration 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. 1.1.2.3 When, in MODES 3, 4, or 5 with the reactor trip breakers closed**,

verify that criticality cannot be.achieved with shutdown group CEA withdrawal 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. 1.2.4 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0X delta k/k at 1'east once per.31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4. 1. 1.2. l.e or 4. 1. 1.2.2.

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.

- The CEA drive system capable of CEA withdrawal.

PALO VERDE UNIT 3 Amendment No. R,69

MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION

3. l. 1.3 The moderator temperature coefficient (MTC);shall, be.within the area of Acceptable Operation specified in the CORE OPERATING LIMITS jfEPORT,.

The maximum uppper 1limit shall be less than or equal to + 0.5 x 10 hK/K/~F for.

a power level of OX RATED THERMAL POWER with a,linear ramp to QbK/K/

I= yt 100K,RATED THERISL POWER.

APPLICABILITY:

MODES 1 and 2*¹.

ACTION:

With the moderator temperature coefficienit outsidle the area of Acceptable Operation,,

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

SURVEI LLANCE REI UIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its '-'limits by confirmatory measurements.

MTC measured values shall be extrapolated and/or.

compensated to permit direct comparison with the above liimitsq

4. 1. 1.3;2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel.

cycl<.'.

b.

C.

Prior to initial operation above. 5X of..RATED THERMAL POWER,'after'ach

'fuel 'iloading.

At any THERMAL POWER, w'ithin 7 EFPD after r'caching a core a'verage exposure of 40 EFPD biurnup into the current cycle.

At any THERMAL POWER; within 7'EFPD after reaching a coreaverage exposure equival,ent to two-thirds of the. expected current cycle end-o'-cyc'le core average burnup.

ef.f

¹See Special Test Excepti,on 3.10.2.

PALO VERDE " UNIT 3 3/4 1-4 AMENDMENT NQ.,42

TABL 3.3-REAC OR ROTECTIVE INSTRUMENTATION I

4 NNNNITTNNNL ITINNT.

I.

TRIP GENERATION A.

Process TOTAL NO.

OF CHANNELS CHANNELS TO TRIP HINIHUH CHANNELS APPLICABLE OPERABLE NODES

~CT ON CA I

Crd 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 8.

Local Power Density - High 9.

ONBR Low B.

Excore Neutron Flux 1.

Variable Overpower Trip 4

4 4/SG 4/SG 4/SG 4

4/SG 4

4 2

3 2 (b) 3 2/SG 3/SG 2/SG 3/SG 2/SG 3/SG 2

3 2/SG 3/SG 2 (c)(d) 3 2 (c)(d) 3 1,

2 1,

2 3*

4*

1, 2

2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ 2¹, 3¹ O

2.

Logarithmic Power Level - High a.

Startup and Operating b.

Shutdown C.

Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators 2 (a)(d) 3 2 (e) 2 (c)(d) 3 1,

2 2¹, 3¹ 3*

4*

5*

9 3, 4, 5

4 1,

2 6,

7 1,2,3*,4*,5*

2¹, 3¹, 7; 10

Il 0

"4 g

3 ~

~ l

TABLE'.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION ACTION STATEMENTS Steam Generator Pressure Low Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF) 4.

Steam Ge

.ator.

Level - Low Steam Generator Level -

Low (RPS)

(Wide Range)

Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF) 5.

Core Protection Cal.culator Local Power Density - High (RPS)

DNBR - Low (RPS)

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

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

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

ACTION 5 With the number of channels OPERABLE one less than required by.the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;

however, the trip breaker associated with the inoperable channel may be closed for up to I hour for surveil'lance testing per Specificati'on 4.3.1.1.

ACTION 6 a.

With one CEAC inoperable, operation may. continue for up to 7 days provided that the requirements Qf Specification 4. 1.3. 1. 1 are met.

After 7 days, operation may continue provided that the conditions of Action Item 6.b are met.

b.

With both CEACs inoperable operation may continue provided that:

PALO VERDE " UNIT 3 3/4 3-7 AMENDMENT NO. 18

II

~

ACTION 7 ACTION 8 ACTION 9 ACTION 10 '-

REAC1 OR PROTECT IVE. INSl'RUMENTATION ACTION

-STATEMENTS'.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .'he DNIBR ma>r gin required by Speoiftica-tion 3;2;4b;(COLS!i in service).or 32.4d (COL$S putj of servi'ce)'is, satisfied arid the Reactor Power Cutback.

System is disabled,

.and 2.

Within, 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

All full-length and part-.'length CEA groups must be wi,thdrawn withi'n the.limi.ts of Specifi'cations 3.'l.'3;5; 3'.1.3.6b, and 3. 1.3,.7b, except during survei.'llaincI -testing pursuant tai the requiremI!nts.of,Speci Ficat ion 4.1.3.]..2.

Sp'eci'fication 3'.1.,3.6b al'lowers CEA-Group 5'nsertion to no further tlhan 127;5 inches.

withdra'wn.

b)'he

'."RSPT/CEAC Inop'erable"'ddressable constant

.in 'the,CPCs

'is..set to indicate that both:CEAC's are inoperable.

c)

The Control Element Drive Mechanism Control, System

{CEDHCS) i

placed ini and subsequently maaintain'edi in the "Standby" mode.exc'ept during CEA'otioni permitte'd"by.SIpecifications

.3i I'.3.5,

3. 1.3;6b an'd'. 3. 1.3;7b, when the CEDHCS may be operatedl in either the

'-'Manubial Gr'up" or "Manual Individual" mode;.

3.

CEA'position surveillance midst meet the requirements of. Specifications

4. 1.3. 1. 1, 4. 1.3.5, 4'.,1.3.6 and

4. 1.3.7 except during surveI11ance testing pursuant to Sjiecification '4. 1'.3.;1.2.

.With 'three or more auto restarts, excluding periodic auto restarts (Code 30 and 'Code 33)" of one'on-bypassed calculator during a '12-hour interval,, demonstrate calculator OPERABILITY by performing a CHANNEL,'FUNCTIONAL'TEST within the nexti

.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the number of'PERABLE charm'els one less than the Hinimum Channeels OPERABLE requ'irememt, restore an inoperable channel tai OII?ERABLf status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open an affected reactor trip breaker within the next h'ouW.

With the number. of OPERABLE channels one less than the Minimum

.Channels OPERABL'E, 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 open the reactor. trip breaker.'ithin

.the iaext liour'.

.Iri'MODES 3; 4, or 5, the Core Protection, Calculator channels ar'e not required to be OPERABLE when the Logarithmic"Power Leve'1' High trip is OPERABLE,w'ith,the trip setpoint lowered to

<10 X of Rated'Thermal

.Power.

.PALO VERDE..UNI'll 3 3/4 3-8 Amendment No.

1B'

~IIII I.I.I EE II EE REACTOR PROTECTIVE INSTRUHENTATION SURVEILLANCE RE VIRESCENT FUNCTIONA

. UNIT D.

Supplementary Protection System Pressurizer Pressure - High II.

RPS LOGIC A.

Hatrix Logic B.

Initiation Logic III. RPS ACTUATION DEVICES CHANNEL CHECK N.A; N.A.

CHANNEL CALIBRATION N.A.

N.A.

CHANNEL NODES IN WHICH FUNCTIONAL SURVEILLANCE

~S

~RE III EE 1,

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

I, 2, 3*, 4*, 5*

B.

Hanual Trip A.

Reactor Trip Breakers N.A.

N.A.

H, R(IO)

I, 2, 3*, 4*, 5*

3*

4*

5*

TliBLE 11~3-i Ci)ntinlladi.

REACTOR PROTIECTIVI= INSTRUMENTATION SURVE][LLANC;E IRE iUIREMENTS

'TABLE NOTAi'IONS

With reactor trip breakers in, the closed position and the CEA drive system ca!parle of CEA wit;hdrawal, and.fuel in the reactor vessel.

(1)

Each SlARTUP or when required with 'thk r'eactor trip breakers closed and the C'EA dr-ive system capable of rod withdrawal, if not pr.'rformed in the previou;s 7 days.

(2)

Heat balance only (CHANNEL FUNCTIlONAL TEST,not inc'luded):

a.

Between 1!5X and SOX of RATED 1IHERMAL POWER, compare the linear power leviel, the CPC delta

'T power and the CPiC nuclear poweri signals to the calorimetric calculation.

If any signal is 'within -0.5X to 10X of the calorimetric then dc'ot calibrate except as required duri,ng initi,al, power ascension after refuel ing.

If'ny signal is less than the calorimetric calculation by more than 0.5X,, then adjust the affected signal(s) to agreie with the caloriimetric calcu'lation.

If'ny signal is greater than the calorimetric calculation by more than 10X then adjust the affected signal(s) to agree with the calorimetric calculation within SX to 10X, b.

At, or above SOX of FIATED THERMAL POWER; compare the 1'inhari pOwer level, the CPC delta T power and the CPC, nuclear power signa'lls.to the calorimetric calculation.

If any.signal differs from,the calorimetric calculaition by an absolute difference of m..r than 2X, then adjust t'e affected sigrral,(s) to agree with the calorimetric calculaition.

During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon r eaching each major testpower.plateau and !prior to proceeding to the next major test power plateau.

(3)

Above 15X of'ATED THERMAL POWER,erify,'that the linear power sub-channel gains of the excore detectors are consistent with the values used to establish the shaipe annealing matrix elements, in the Core Protect,ion Calculators.

(4)

Neutron dietectors may-be excluded from CHANNEL CALIBRATION.

(5)

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

POWER, thie incore detectors shal',l be used to determine the shape i

annealing matrix elements and the Core Protection Calculators shall use these elements.

PALO VERDE UNIT 3 3~r4 3-Ie Amendment No. KM&,69

TABLE 4.3-1 Continued REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREHENTS TABLE NOTATIONS (6)

(7)

(8)

This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into the channel:

as close to the.sensors as practicable to verify.OPERABILITY including alarm and/or trip functions.

Above 70X of RATED THERMAL POWER, verify that the total steady-state RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation or by calorimetric calculations and if necessary, adjust the CPC addressable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate.

The flow measurement uncertainty may be included in the BERRl term in the CPC and is equal to or greater than 4X.

Above 70X of RATED THERMAL POWER, verify that the total steady-state RCS flow'rate as indicated by each CPC is,less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure.

instrumentation and the ultrasonic flow meter adjusted pump curves or calorimetric calculations.

The quarterly CHANNEL'UNCTIONAL TEST shall include verification that the correct (current) values of addressable constants are installed in each OPERABLE CPC.

(10)

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

PALO VERDE UNIT 3 3/4 3-16a Amendment No.

69

0 il s

g Vs

~

J

3 4.9 REFU ING OPERATIONS 3 4.9. 1 BORON, CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head

removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be main'tained uniform and within -,the limi.t specified in the Core Operating Limits Report,(COL'R).

I TT:

OOOO 6, ACTION:

With the requirements of the above specification not satisfied',

immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate,and continue boration at greater than or equal to 26 gpm of a solution containing

> 4000 ppm boron. or its equivalent until the boron concentration is within limits.

'SURVEILLANCE RE UIREMENTS 4.9. 1. 1 The boron concentration shall; be determined to be within the limit specified in the COLR prior to:

a.

Removing or unbol.ting the reactor vessel

head, and b.

Withdrawal of any full-length CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and.the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The reactor shall be maintained in NODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.

PALO VERDE UNIT 3 3/4'-1 Amendment No.

69

REFUELING OPERATIONS 3/4. 9.2 INSTRlNENTATION LINITING CONDITION FOR OPERATIONI 3.9 2

As a minimum, two styrtup channel neutron flux monitors shall be OPERABLE and operatingeach with.continuous visual indication in the contro11 room and one with audible indication in t,he coIntlinheAt and control room.

APPLICABILITY" MODE 6.

ACTMN:

a.

With one of the above-.required monitors inoperable or not operating<

immediately suspend all operations,invollving CORE ALTERATIONS or pos'itive reactivity changesl.

With both o'f the above required monitor.

inoperable or not operating, determine the boron concentration, of the Reactor Coolant Sys'em at, least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVK1LLANCE R~E ijIREMENTS

4. 9 2 Each startup channel neutron.f'lux monitor shall be demonstrated OPERABLE by performance ot:

a.

A CHANISEL CHECK at least onice'per 12 'hours',

b.

A CHANNEL 'FUNCTIONAL l'EST Within 8'ho'urs prior to the initial.'tart of CORIE A,LTERATIONS, and c.

A CHANNEL FUNCTIONAL lEST ait )ea'st once per 7 days.

%AM VERDE -

IIJHIT 3 3/4 9-2

3 4. 1 REACTIVITY CONTROL SYSTEMS BASES 3 4. 1. 1 BORATION CONTROL 3 4.1.1.1 and 3 4.1.1.2 SHUTDOWN MARGIN AND K The. function of SHUTDOWN MARGIN is to ensure that the.reactor remains subcritical following a design, basis: accident or:anticipated"operational occurrence.

The function of K, is to maintain sufficient subcriticality to preclude inadvertent criticality following ejection of a single control element assembly (CEA)..

During operation in MODES 1 and 2, with k << greater than or equal -to 1.0, the.transient insertion limits of Specification. 3. 1.3.6 ensure that, sufficient SHUTDOWN MARGIN is available.

SHUTDOWN MARGIN is the amount by which the core is subcritical, or would be subcritical immediately following a reactor trip, considering a single malfunction resulting in the.highest worth CEA failing to insert.

K, is a

measure of the core's reactivity, considering a single malfunction resulting in the highest worth inserted CEA being ejected..

SHUTDOWN'ARGIN,requirements vary throughout the core life as a 'function of fuel depletion and reactor coolant system (RCS) cold leg temperature (T,~).

The most restricti.ve condition occurs at EOL, with T<< at no-load'perating temperature, and is associated

-with a postulated steam line break accident and. the resulting uncontrol-led RCS cooldown.

In the analysis oF this acci'dent, the specifi.ed SHUTDOWN MARGIN is required'o control the reactivity transient and ensure that the fuel performanc'e and offsite dose criteria are satisfied.

As (initial) T<< decreases, the potential RCS cooldown and the resulting reactivity transient are less severe and, therefore, the required SHUTDOWN MARGIN also decreases.

Below T<< of about 350 F,, the inadvertent deboration event becomes limiting,with respect to the SHUTDOW:.'ARGIN requirements.

Below, 350'F, the specified'HUTDOWN MARGIN ensures that sufficient time for operator actions exi'sts between the initi'al indication of.

the.deboration and: the'otal loss of shutdown margin.

"Accordingly, with, the reactor trip breakers closed and..the CEA drive system capable of CEA withdrawal, the SHUTDOWN MARGIN'equirements are based upon these l,imiting conditions.

Additional events considered in establishing requirements on SHUTDOWN MARGIN that are not limiting with respect to the Specification limits are single CEA withdrawal and startup oF'n inactive reactor coolant pump.

K, requirements vary with the amount of positive reactivity that would be introduced assuming the CEA with the highest inserted worth ejects from the core.

In the analysis of the CEA ejection event, the K, requirement ensures that the radially averaged enthalpy acceptance criterion is satisfied, considering power redistribution. effects.

Above T<< of 500 F, Doppler reactivity feedback is sufficient to preclude the need for a specific K, requirement.

With all CEAs fully inserted, K. and SHUTDOWN MARGIN requirements are equivalent, in terms of minimum acceptable core boron concentration.

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

REACTIVITY CONTROL SYSTEMS BASES SHUTDOWN MARGIN AND ~K.,> {continued)

The requirement prohibiting.cr'itiicality due to, shutdown group CEA,moyemeni; is associated with the assumptioiis used in Ithp ana)ysis,,of.uncontrolled CgA,,

withdrawal from subcriti'cal. conditions.

Due

'$o jthe high differentia'l reactivity worth,of the. shutdown CEA groups, thej atialysjis assumes that, the

,initial,shutdown reactivity is, such.that the reactor wi'Ill ren>aiin subcritigal in the event of, unexpected or uncontrolled shutdown: gr'up withdrawal.

Other technical specifications that referepc'e; the,!ipecifications,on SKUT-DOWN MARGIN.or K.-, are:

.3/4.'1.2,

'BORATION, SYSTEMS,, 3/4. 1-.3, MOVABLE

'CONTROL ASSEMBLIES 3/4

~ 5 ~ 1',

REFUELING OIPERATIONS'ORON CONCENTRATION',3/4, 1O. 1

~ SHUT-DOWN MARGIN AND K..

CEA WORTH TESTS, and,3/4.10.9,.

SHUTDOWN HARI3IN AND.

K). CEDHS TESTIMi.

3 Ac1.

1 t3

.NOOERATOII TENIAERATIIRE COEFI=TC~IEIIT 'TC The limitations on moderator temp'erature, coeffic'ient (HTC) tare provided to ensure that thIe assumptions used in the accident and transient analysis remain valid; through--each fuel cycle.

, The,surveil'1'ance r'equirements for measurement of the.HTC during, each fuel cycle are adequate 'to confirmthe.HTC value since, this coefficiient chaing'es,,slowly due principally to the reductioki in RCS boron concentrati6n associated with fuel'.burntup'.

Th'e.confirmation that

,the measured HTC, value, i within its'limitlproyides. assur'ances that.tthe coeffi:-

cient will be maintained,,w'ithin acceptable, values Ithroughout 'each fuel'ycle; 3 4.1..4

. MINIMUM. IEHPEflATURE-FOR CRITICALITY This'f3eci,fication ensiires that,the reactor, will-not. be made critica'1 with the Reactor Coolant System cold 'leg. temperature less, than ',545'.F.

This limita-tion is required to ensure (1) themoderator temperature coe'fficient is within its,.analyzed; temperatu're

range, (2).the pr'6tectivet instrumentation, is within

,its normal operatirig range,

-(3).a iriinimum.temperature, is provided for Stpecial Test Exception,3/4.10.4,

'and (4) the reactor

.~egseil is above.its minimilm RTgQ7 temperature.

PALO VERDE -..'UNIT 3' 3/4 1-la Amendment No. 49, 69

~

~.Q 3 4.9 REFU LING OPERATIONS BASES 3 4 9.1 BORON CONCENTRATION The limitations.on reactivi,ty condi,tions during REFUELING ensure that:

(I).the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform boron concentration

.is maintained.for reactivity control in the water volume having direct access to the reactor vessel.

These limitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.

The boron concentration limi't specified in the COLR is based on core reactivity at the.beginning of each cycle (the end of refueling) with all CEAs withdrawn and. includes an uncertainty allowance.

This boron concentration limit will ensure a K,<<.of g0.95 during the refueling operation.

3 4.9.2 INSTRUMENTATION The OPERABILITY of the startup channel neutron.flux monitors ensures that redundant monitoring capability,is available to detect changes in the reactivity condition of,the core.

3 93 C

Y IME The minimum requirement for reactor subcriticality.prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to al.low the radioactive decay. of the short lived fission products.

This decay time is consistent, with the assumptions used in the safety analyses.

3 4.9.4 CON AIN ENT BUILDING PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.

The 'OPERABILITY and closure restrictions are sufficient. to restrict radioactive material release from a fuel element rupture based upon the.lack of containment pressurization potential. while in the REFUELING MODE.

3 4.9.5 COMMUNICAT ONS The requirement for communications capability ensures that refueling, station personnel'an be promptly informed of significant, changes in the facility, status or core reactivi'ty condition during CORE ALTERATIONS.

PALO VERDE UNIT 3 B 3/4 9-1 Amendment No. 69

0 REFUELING OPERATIONIS

'BASES 3/4. 9. 6 REFUELING MACHINE The OPERABII ITY requirements for the refuelir'ig inachine ensure that:

(1) the machine will be used for movement of fuel assemblies, (2) the imachine

.has sufficient load capacity to-lift a fueil.assembly,.

and (3) 'the 'coi'e internals and-pressure ves. el are lprotected from.excessive.lifting force in the event'hey are inadvertently erigaged dlur.ing.lifting operations.

3/4.'9.:7 CRANE TRAVEL -. SPENT FUEL STORAGE POOL BUILDING The restriction iin movement of loads in. exceeds of 'the nominal, weight,. of a fuel. assembly, CEA a'nd ~associated

'handling thol ov'er other'uel, assemblies in-'the storage pool.ensures that ini the event'his'oad, is dropped (1) the activity release will'e limitedl to'. that. contained in a single fuel assembly, and- (2) any possiibl'e clistortion of. 'fuel in the storage racks will not..reSult in a critical array.

This assumpt',ion is con0istent with the activity release assumed in the safety 'analyses.

'~T. 'I'LT!

The requirement that. at 'ileast one shutdown cooling loop be.'in o'peration, and circulating -reactor coolant at a flow late equal to or greatei than 3400 gpm (actual) ensure.

that (1) sufficient cooling capaci'ty is, available to remove decay

heat, and maintai'n the water iii the reactor pressure

-vesse'I below 135'F as required during the REFUELING MODE, (2) sufficierit:coo'lant circulation is maintained through the reactor core to min'imize the effects of a,boron dilution incident ar>d'prevent boron'trat'i1I'!i'cation, and (3):the hT'acr()ss the core will'be maintaine'd! at less than 75,F'.during the-REFUEL'ING MODE..

The required flowr'ate of > 3400 gpim (a'ctual) ensures that; at 288 hours0.00333 days 0.08 hours 4.761905e-4 weeks 1.09584e-4 months after reactor shu'tdown suffTcient cooling capacity is available'o remov'e.decay heat and maintain the water in the reactor pressur'e vessel'elow 135 F as re.'overed during REFL'ELING MODE; this assumes,-a, shutdown.cool.ing heat exchanger cooling water flowrate of 14000

gpm, a cooling. water inlet temperature of < 105 F at, 27 1/2, hours after reactor

shutdown,

-cond the, decay 'heat curve. of.CESSAR-!F~

Figure 6.2.1-1 and reactor operation for-trio years 'at 4000 MWt.

1he 3780,'gm..'.

in the specification 'includes alii instrument uncertainties

.ir>eluding. the. 300 F'alibration temperat,ure of th'e.f1lowtransmitters.

Without a. shutdlown coo'ling'rain in operation steam,may be generated; therefore, the containment

.'hould be sealed off, to,prevent.e.'cape of arly radioactivity, and.any operations that would cause'n increase in deca/ heat should, be secured,.

The requirement, to.have two shutclown cooling 'loops OPERABLE when there is less than 23 feet of water above the r'eactor pressure vessel

flange, erisui ed

.that a single fai'lure,'<)f the operating shutdown cooling loop will not r'esult, in a complete los's of iiecay heat removal capability.

'With the reactor 'verse,l' head removed and P3 feet of water -a'bove the reactor pressure vessel:flange,

~a large heat sink is available for core cooling, thu. in 1.he event.of a failure of the operating shutdown cooling loop, adelquhtel time is pr;ovided, to -initiiate emergency procedures to cool'he.core.

PALD VERDE - UNIT 3 iB 3/4 9-l2' AMENDMENT NO. 33

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core 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:

a ~

b..

C.

d.

e.f.

g h.l.

J

~

k.

Shutdown Margin 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 Regul'ating 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 Tq.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'pproved by the NRC in:

'a ~

b.

c ~

d.

"CE Method for Control Element Assembly Ejection Analysis,"

CENPD-0190-A, January 1976 (Hethodology 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. 1.2, 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 (Ho 'e 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. l. 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 Tq).

"Hodifi'ed Statistical Combination of Uncertainties,"

CEN-356(V)-P-A Revision Ol-P-A, Hay 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 3 6-20a Amendment No. 4~,

69

e.

"Calculative Methods for the CF. Liarcie. Break LOCA Evaluation Model i'r the Analysis of CE and W DesigAed NSSSj".CENPD-132, 'Supplement '3-P-A,,

June 1985 (Hiethodology for Specification 3.2. 1, Linear Heat Rate).

"Calculative Methods fair the CE Sinai'1 Bre'ak'LOCA Evaluation'odel,"

CENPD-137-P, August 1974 (Hethodo'logy for Specificatio'n h.2.1, Linear Heat Rate).

9 ~

h.

"Calculative Methods for the CE Small Break LOCA Evaluation'a'del,"'ENPD-137-P, Supplement lf', 3anuary 1977 (Methodology 'for Specification 3.2.1, Linear Hekt Aatk).

Letter.,:

0.

D. Parr (NRC) to F.'.',

S'tern (CE), datecl June 13, 1975 (NRC Staff Reviiew.of the Combus'tion Engineering ECCS'valuation Model).

NRC approval for:

6.9;1;,10f..

Letter:

K. Kniel (NRC) tai A.

E'. SchIere'r (CE'),, dated'Septiemberi 2N, 1977 (Evaluation of Topi'cal Reports CENPD-133,. SupplemI.nt 3-P ancl CENPD-137, Supp'lement,l-P).

NRC approval.for 6.'9.1.10.g.

The core operating limits shall be determined so that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraul,'ic l.imits, ECCS limits, nuclear limits such as shutdown margin and transient ancl analysis limits) of the saf'ety,analysis are met.

-The CORE OPERATING, LIMITS REPORT,, includikg Ln).mid-cycle revisions or supplements

thereto, shall be provided upon iisduahce, for each reload cycle, to the NRC Document Con'trol'Desk with copies td the 'Regiohal Admini. trator.

and Resident Inspector.

PALO VERDE - VNIT 3 6-20b AMENDMENT NO.. XR, 55

ML17311B166

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