ML17312A801

Amends 108,100 & 80 to Licenses NPF-41,NPF-51 & NPF-74, Respectively,Revising TS 1.26 to Increase Authorized Rated Thermal Power

ML17312A801
Person / Time
Site:	Palo Verde
Issue date:	05/23/1996
From:	Russell W NRC (Affiliation Not Assigned)
To:
Shared Package
ML17312A802	List: ML17312A801 ML17312A803
References
NUDOCS 9606110095
Download: ML17312A801 (62)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 R

C 0

Y ET T NO.

1 Amendment No. 108 License No. NPF-41 The Nuclear Regulatory Commission (the Commission) has found that:

A.

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

Power, and Southern California Public Power Authority dated January 5,

1996, as supplemented by letters dated April'9, 1996, Hay 1,

1996, Hay 10, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

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

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

The issuance of this amendment is in accordance wi'th 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 paragraphs 2.C(l) and 2.C(2) of Facility Operating License No.

NPF-41 are hereby amended to read as follows:

'Pb0bii0095 '&0523 PDR ADOCK 05000528 P

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(2)

Arizona Public Service Company (APS) is authorized to operate the facility at reactor core power levels not in excess of 3876 megawatts thermal (lOOX power) in accordance with the conditions specified herein and in Attachment I to this license.

The items identified in Attachment I to this license shall be completed as specified.

Attachment I is hereby incorporated i'nto this license.

c on Pl n

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

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

Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of successful completion of the Cycle 7 reload analysis and to be implemented prior to startup from Unit I refueling outage six.

FOR THE NUCLEAR REGULATORY COHNISSION William T. Russell, Director Office of Nuclear Reactor Regulation Attachments:

1.

Page 4 of License 2.

Changes to the Technical Specifications Date of Issuance:

May 23, 1996 Page 4 is attached, for convenience, for the composite license to reflect this change.

Please remove page 4 of the existing license and replace with the attached page.

41 b

~

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and order s of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

Arizona Public Service Company '(APS) is authorized to operate the facility at reactor core power levels not in excess of 3876 megawatts thermal (100X power) in accordance with the conditions specified herein and in Attachment I to.this license.

The items identified.in Attachment I to this l.icense shall be completed as specified.

Attachment I is hereby incorporated into this license.

(2) echnical S

e 'ficati ns nd nvironment 1 Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 108, 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) n itio s This license is subject to the antitrust conditions delineated in Appendix C to this license.

(4) e t ff eri R

uire t

APS shall have operators on each shift who meet the requirements described in Attachment 2.

Attachment 2 is hereby incorporated into this license.

(5) o

1-d'n'ti 1

e P

o e tion 4

SER nd Any changes in the, Initial Test Program described in Section 14 of the FSARs (Palo Verde and CESSAR) made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

• The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or.its supplements wherein the license condition is. discussed.

Amendment No. 108

4l I

108 S

NDM NT G

C NO ST 50-8 0

-4 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

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

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

INSERT 1-5 3/4 1-'5 3/4 1-19 3/4 2-8 3/4 4-7 3/4 4-8 B 3/4 1-5 B 3/4 4-12 1-5 3/4 1-5 3/4 1-19 3/4 2-8 3/4 4-7 3/4 4-8 B 3/4 1-5 B 3/4 4-12

DF ITO 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the

FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

P AN K

G ACTOR F 1.22 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal, plane, excluding the effects of azimuthal tilt.

1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a.nonisolable fault in,a Reactor Coolant System component body, pipe wall, or vessel wall.

P 0 POG P

1.24 The PROCESS CONTROL PROGRAH (PCP) shal,l contain the current formulas,

sampling, analyses,

test, and determinations to be made to ensure that processsing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid. radioactive waste.

U 1.25 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature,

pressure, humidity, concentration,.

or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1.26 RATED THERHAL POWER shall be a total reactor core heat transfer, rate to the reactor coolant of 3876 HWt.

1.27 The REACTOR TRIP SYSTEH

RESPONSE

TIHE shall be the time interval from when.the monitored parameter exceeds its trip setpoint. at the channel sensor until electrical power is interrupted to the CEA drive mechanism.

PALO VERDE UNIT 1 1-5 Amendment No. 27-,M,108

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CO TRO SYST S

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. ON 3.'1. 1'.4 The Reactor:Coolant System lowest operating loop temperature (T,~)

shall be greater than or equal'o 545 F.

NODES 1 and 2P.

KIIQH:

With a Reactor Coolant System operating loop temperature (T ~) less than 545'F, restore T

~ to within its limit within 15 minutes or be in 'HOT STANDBY within the next 5 minutes.

SURVEI LAN UI EH NTS

4. 1. 1.4 The Reactor Coolant System temperature (T,~) shall be determined to be greater than or equal to 545 F:

a ~

b.

Within 15 minutes prior to achieving, reactor criticality, and At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T<< is less than 550 F.

PWith K,<< greater than or equal to 1.0.

PALO VERDE UNIT 1 3/4 1-5 Amendment No. 54-,69-,T7-, 108

REACTIVITY C t'ROL SYSTB5 3/4.1.2

.BORATION SYSTENS FLOW PATHS -

SHUTDOWN LINITIKG CONDITION FOR OPERATION 3.1.2.1 As a iainie~e, one of the fo'lldvfhg bor'on 'in)ection flow +th's Shall be OPERABLE:

a.

If only the spent fuel pool in Specif'tcatfon.3.1.2.5a.

fs OPERABlZ,

, a flow path frow tihe spent fuel poc>1 via a gravity feed connection and a charging puap to the Reactor Coolant Systee.

b.

If only the refueling water tank in Specification 3,.1.2.5b.,is

OPEIVBLE, a flow path -from'tQ rief&ling'water tank via Wither a charging pump, a h;igh presSure safety infection puap, or ta low pres-'ure safety injection puap 'to'thi Reactor Coolant Systea.

iPPLICABILITY:

HODES 5 and 6.

ACTION:

Nth none of the above flew paths OPERABLE,'iuspend all operations iin4olking CORE ALTERATIONS or pos'itive reactivi'ty 'changes.

SURVEILLANCE REI~UIRENENTS 4.1.2.1 At 1'east one of the above required f1low paths shall be dea~nstrated OPERABLE at least once per 31 days by verifying that each valve (aanual, powir operated, or automatic) in the flow path that is not locked, healed, or otherwise secured in position, is in its correct position.

PALO VERDE - UNIT 1 3/4 li6'NENDNENTNO. 69

A T C

D LIMIT G

CO ION 0

OPE TION

3. 1.3.4 The individual full-length (shutdown and regulating)

CEA drop time, from a fully withdrawn position, shall be less than or equal to 4 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90X insertion position with:

t.

'a ~

T,o,d greater than or equal to 550'F, and b.

All reactor coolant pumps operating.

~APP P

~ACT ON:

~

MODES 1 and 2.

a ~

With the drop time of any full-length CEA determined

.to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to MODE 1 or 2.

URV A

UIREM NTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a.

For all CEAs following each removal and reinstallation of the reactor vessel

head, b.

c ~

For specifically affected, individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific

CEAs, and At least once per 18 months.

PALO VERDE UNIT'/4 1-19 Amendment No. 69,i08

REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INSERTION LIMIT II LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdra'wn'to at least 144.75 inches.

APPLICABILITY:

MODES 1 and 2">F.

ACTION:

,E With a maximum of one shutdown CEA withdraw'n to less than 144.75 inches, except for surveillance testing pursuant to Spec)fication-4.1.3.1.2, within l,hour either,:

a.

Withdraw the CEA to at least 144.75 inches, or b.

Declare the CEA inoperable and ccinply with Specification 3.1.3.1.

RUR ERE R

E REI UERE E E 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least 144.75 inches:

a.

Within 15 minutes prior to withdrawal of any CEAs in regulating

,groups during avi approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter except during time intervals when both CEAC's are inoperable, then verify the individual CEA position,s an't,')eaist once.per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

See Special Test 'lException 3.10.2.

Alith K ff greater than or equatl to 1.

'PALO VERDE - UNIT 1 3/4 1-20 AN~NCN~NT NO. ]IN, 69

POMER DISTRIBUTION LIHITS 3/4.2. 6 REACTOR COOLANT COLD LEG TEMPERATURE LIHITIKG'ONDITIONFOR OPERATION 3.2.6 The reactor coolant cold leg teaperature (T ) shall be vfthfn the Area of Acceptable Operation shown fn Figure 3.2-1.

APPLICABILITY:

HODE 1" and 2*8.

ACTIDN:

fifth the reactor coolant cold leg teaperature exceedfng fts lfaft, restore the teeperature to within fts lfaft wfthfn 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be fn HOT STANDBY wfthfn the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREHENTS 4.2.6 The reactor coolant cold leg temperature shall be determfned to be wfthin its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"See Special Test Exception 3.10.4.

NNth K << greater than or equal to 1 PALO VERDE - UNIT 1 3/4 2-7 AHENDHENT NO. f7, 69

575 570 O

W 565 P-560 555 550 0

O, 545 570 56S (l00,560) 560 550,'i45

'540 10 20 30 40 50 i

i,60 i

i,70 80 90 100 CORE POWER LEVEL,% OF RATED THERMALPOWER (3876 MW)

REAC1OR COOLANT COLD LEG TEMPERATIJRE! vsCORE POWER LEVEL F'IGURE 3.2-1 REACTl)R COOLANT COLD LEG TEHPERATUREI V$.

,'CORE

'POMER LEVEL PALO VERDE UNIT 1

3/4 2-8 Amendment No. PJ-,77, lO8

L HIT G

CO TION 0

OPE TION 3.4.2. 1 A minimum of one pressurizer code safety valve shal.l be OPERABLE with a lift setting of 2475 psia +3, -1X*.

NODE 4

~ACTI a ~

b.

Mith no pressurizer code safety valve OPERABLE, immedi'ately suspend all operations

.involving positive reactivity'hanges and place an OPERABLE shutdown cool:ing loop into operation.

The provisions of Specification 3.0.4 may be suspended for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for entering into and during operation in NODE 4 for purposes of setting the pressurizer code safety valves under ambient (HOT) conditions provided a preliminary cold setting was made prior to

,heatup.

SURV ILLANCE R U

R E T 4.4.2. 1 No additional Surveillance Requirements:other than those required by Specification 4.0.5.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

PALO VERDE UNIT 1 3/4 4-7 Amendment No. 87-;75, 108

6ILHR ILI'2IJU'E.QI

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3.4.2.2 All pressurizer code safety valves shall be OPIERABLE.with

a. lift setting'of 2475:. psia +3,

.IX~.

Mith one pressurizer code. sifety valve inotierable, either restore the inoperable valve to OPIERABLIE status within 15 minutes, or be 'in at least HOT STANDBY within 6 Ihours.and. 'jn H01'HUTDOWN within the, following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with

.the shutdown cooling system suet'ion line reli'ref valve.s aligned to'providei overpr essure protection f'r'he Reactor Coolant System.

s~vliL tJllgi&T::

4.'4.2.2 No additional, Surveillance Requiremehts other than those required by

.Specification '4.0."5.

• The lift setting pressure shaill,correspond to ambient conditions of the valve at nominal operating temperature arid.-pressure PALO VERDE UNIT 1 3/4, 4-8 Amendment,No. M

,75, 108

and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of at least two CEA position indicator channels is required

, to determine CEA positions and.thereby ensure compliance with the CEA alignment and insertion l.imits.

The CEA "Full In" and "Full Out" limits provide an additional. independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

The maximum, CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses.

Measurement with T,~ greater than or equal to 550 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Several design steps were employed to accommodate the possible CEA guide tube wear which could arise from CEA vibrations when fully withdrawn.

'Specifically, a programmed insertion schedule will be used to cycle the CEAs between the full out position ("FULL OUT" LIMIT) and 3.0 inches inserted over the fuel cycle.

This cycling will distribute the possible guide tube wear over a larger area, thus minimizing any effects.

To accommodate this programmed insertion schedule, the fully withdrawn position was redefined, in some cases, to be 144.75 inches or greater.

The establishment of. LSSS and LCOs requires that the expected long-and short-term behavior of the radial peaking factor s be determined.

The long-term behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount of CEA insertion

assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle.

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

The magnitudes of such perturbations depend upon the.expected use of the CEAs during anticipated, power reductions PALO VERDE UNIT I B 3'/4 1-5 Amendment No. 108

0 REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES~iContinue+d and load maneuvering.,

Analyses are performed based oh t'e expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyses.

CEA insertions are determined and a consistent set of radial peaking factors defined.

The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the assumed mode of operation usedin the 'analyses and prov'ide a means of preserving -the assumpt'ions on CEA insertions used.

The limits spe'ci-fied serve to limit the behavior of the radial peaking factors within the bourids determined from analysi,s.

The actions specified serve to limit the extent o)F radial xenon redistri'bution effects to those accommodated in the analyses.

The Long and Short Term'nsertion Limits of'pecif'ications 3,.1.3. 6 and 3.1. 3. 7 are specified +or the plant which has been designed for primarily base loaded opera-tion but which has the ability to accommodate a limited amount, of load maneuvering.

The Transient Inseirtion Limits of Speciifications 3.1.3.6 and 3.1.3.~7 and the Shutdown CEA Insertion Limits of Speciftcation 3.1.3..5 en.ure that (1) the minimum SHUTDOWN MARGIN is mainta',ined, and (2) the potential effects. of a CEA ejection accident are limited to,acceptable levels.

Long-term operation at the'ransient Insertion Limits i. not permitted since such oi)eration could have effects on the core power di. tribution which ciould i'nvalidate assumptions used

'o determine the behavior of -the radial peaking factors.

The PVNGS CPC and COLSS systems ar'e responsible for the.safety and mohitor$ ng functions, respectively,, of the reactor core.

'GLSS m'onito'rs the DNB Power Operating Limit (POL) at>d various operating parameters to helt~ the operdtot main-tain'lant operation within the limiting conditions'or ope'ration (LCO).',Operat-ing within the LCO guarantees that in the event, of an Anticipated Operat'iohal Occurrence (A00), the CPCs will;ps+vide a r6actoH trip iri time to prevent un-'cceptable fuel damage.

The COLSS reserves the Requi'red Overpower'argin (RGPM) to. account for the Loss of Fl'ow (LOF) and CEA'isoperation transients.

When the COLSS is Gut of Service (COOS), the monitoring function is performed via the CPC calculcLtion of DNBR in conjunction with Technica11 Specification COOS Limit, Lines specified in.'he CORE OPERATING LIMITS REPORT which restrict the reactor power suffiCie'ntly to.preserve the ROPM.

The reduction of'he CEA deviiation penalties, in accordance with the CEAC (Control Element Assemb'ly Ca'Iculator) sensitivity reduction program has been performed.

This task. involved setting many of ttie inward single CEA deviation penalty factors to 1.0; An inward CEA deviation event iin effect would "not,be accompanied by the application of the CEA deviati'on penalty in either the CPC DNB and LHR (Liriear Heat Rate) calculations for.those CEAs with the reduced penalty factors.

'The protection

)For an inward CEA deviation event is thus accounted for separately'.

PALO VERDE - UNIT 1 B 3/4 1"6 AMENDMENT NO.

51, 6g

BASES TS C ntinu d 5 EFPY, etc.

and are based upon the irradiation damage prediction by the end of the period.

Accordingly, each time P-T limits change, the LTOP system needs to 'be re-analyzed and modified, if necessary, to continue its function.

A typical LTOP system includes pressure relieving devices and a number of administrative and operational controls.

Each of the, Palo Verde Units has a similar LTOP system that includes two Shutdown Cooling System suction line relief valves for transient mitigation.

Each relief valve has an opening setpoint of 467 psig which, in combination with certain other limiting conditions for operation contained in Technical Specifications, comprises the LTOP system.

Previously, the LTOP enable temperatures during heatup and cooldown have been determined at the intersections between a horizontal line corresponding to the safety valve setpoint (2475 psia). and the most limiting P.-T limit curves for heatup and cooldown, respectively.

Note that the enable temperature generally identifies the upper temperature limit below which the LTOP system has to be operable.

In this analysis, the LTOP enable temperatures were determined in accordance with a definition contained in the latest revision of the Standard Review Plan 5.2.2.

According to SRP 5.2.2 the LTOP enable temperature is "the water temperature corresponding to a metal temperature of at least RT >> +

90'F at the beltline location (1/4T or 3/4T) that is controlling in tFe Appendix G limit calculations."

The heatup and cooldown rate limitations assure, the limits of Appendix G to 10 CFR 50 will not be exceeded with overpressure protection provided by the primary safety valves.

The various categories of load cycles used for design purposes are;provided in Chapters 3

and' of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are l,imited so as not to exceed the limit lines of Figures 3.4-2a and 3.4-2b.

This ensures that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

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

These programs are in accordance with Secti'on XI of the ASHE.Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where s'pecific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASIDE Boiler and Pressure Vessel

Code, 1974 Edition and Addenda through Summer 1975.

PALO VERDE UNIT 1 B 3/4 4-12 Amendment No. SR,l08

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gA8 REGS

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

DOCKET NO.

STN 50-529 PALO VERDE NUCLEAR'ENERATING STATION UNIT NO.

2 n

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

100 License No.

NPF-51 r

I 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 January

.5,

1996, as supplemented by letters dated April 19,

1996, May 1,

1996, and May 10, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the 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; 2.

D.

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

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

Accordingly, the license is, amended by changes to the Technical Specifications

.as indicated in the attachment to this license amendment, and paragraphs 2.C(1) and 2.C(2) of Facility Operating License No.

NPF-51 are hereby amended to read as follows:

~!

4

3.

(1)

Maximum Power Level Arizona Public Service Company,(APS) is authorized to operate the facility at reactor core power levels not in excess of 3876 megawatts thermal (IOOX power) in accordance with the conditions specified herein and in Attachment 1 to this license.

The items identified in Attachment 1 to this license shall be completed as specified.

Attachment 1 is hereby incorporated into this license.

(2)

Technic l S ecific tions and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

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

APS shall operate the facility in accord'ance with the Technical Specifications and the Environmental Protection

Plan, except.where otherwise stated in specific license conditions.

This, license amendment is effecti.ve as of its date of issuance to be implemented within 30 days of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION William T. Russell, Director Office of Nuclear Reactor Regulation Attachments:

1.

Page 4 of License 2.

Changes to the Technical Specifications Date of Issuance:

May 23, 1996 Page 4 is attached, for convenience, for the composite license to reflect this change.

Please remove page 4 of the existing license and replace with the attached page.

1

~

C.

This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated.below:

Poer

'ev Arizona Public Service Company (APS) i s authorized to operate the facility at reactor core power levels not in excess of 3876 megawatts thermal (lOOX power) in accordance with the conditions specified herein and in Attachment 1 to this license.

The items identified in Attachment 1 to this license shall be completed as specified.

.Attachment 1 is hereby incorporated into this license.

(2) c ifi tio s d

nvir nme t 1 Protection Plan The Technical Specifications -contained in Appendix A, as revised through Amendment No.10a, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated into this lice'nse.

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

Plan, except. where otherwise stated in specific license conditions.

(3)

Cni'io This license is subject to the antitrust conditions delineated in Appendix C to this license.

(4) er t St f x eri ce Re uirem nts Section

13. 1.2 SSER (5)

APS shall have a licensed senior. operator on each, shift who has had at, least six months of hot operating experience on the same type of plant,.including startup and shutdown experience and at least six weeks at power levels greater than 20X of full power.

r ton 4

S S

Any changes in the initial test program described in Section 14 of the 'FSARs (Palo Verde and CESSAR),

made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b)'ithin one month of such change.

• The parenthetical notation following the title of many license conditions, denotes the section of the Safety Evaluation Report and/or its supplements

,wherein the license condition is discussed.

Amendment No.~<0

Cl

~

T 0

G C

NO.

NPF-51 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

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

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

~SJLT 1-5 3/4 1-5 3/4 1-19 3/4 2-8 B 3/4 1-5 1-5 3/4 1-5 3/4 1-19 3/4 2-8 B 3/4 1-5

0 4

41

D I

1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear. characteristics, of the reactor core and related instrumentation and (1) described in Chapter 14.'0 of the

FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

Fxy 1.22 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to,plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

1.23 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a: Reactor Coolant System component body, pipe wal.l, or vessel wall..

1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,

sampling, analyses,

test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61,.and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid: radioactive waste.

PU G

1.25 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure,,

humidity, concentration, or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

1.26 RATED THERMAL POWER shall be a total reactor core heat transfer rate to

.the reactor coolant of 3876 MWt.

1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor

,.until electrical power is interrupted to the CEA drive mechanism.

PALO VERDE UNIT 2 1-5 Amendment No. 43-;48, 100

II 4~

~"

R CAL' GC 0

0 0

TO

3. 1.I1.4 The Reactor Coolant.System lowest operating loop temperature (T<<)

shall be greater than, or equal to 545'F..

NODES 1 and 2t.

JKIEH Mith a Reactor Coolant System operating loop temperature (T,d) less than

'545 F, restore T to within its limit within 15 minutes or, be in HOT STANDBY within tfPe next 15 minutes..

SURVEI LA CE UI EVENTS

4. 1. 1.4 The Reactor Coolant System temperature (T<<) shall be determined to be greater than: or equal to 545'F:

a.

Within 15. minutes prior to achieving reactor criti'cality, and b.

At least once per 30 minutes when the reactor, is critical and the Reactor Coolant 'System T<< is less than 550'F.

With K,<< greater than or equal to 1.0.

PALO VERDE UNIT 2 3/4 1-5 Amendment No. 39-,S~, 100

3/4; 1.2'0 ON SYS7'EM il

~FLtN P THS

-. SklrrDOW LIMITING'CONDITION IVOR OPERATION

'.1.2.1 As a ainimum, one of the following,boron injection flow p'aths shall be OPERABLE;:

,a.

If only thi,spent fuel pool in Specification 3.1.2;5a..is

OPERABLE, a flow path, from th'e spent fuel pool via a gravity feed Connection and a charging pump te.the Reactor Coolant Systea.

b.

If only the refueling water tank in Specification 3.1.2.5b. is

OPESSLIE, a, flow path from the refuel,lng.water tank v'$a either' charging pump, ii high pres. ure safety in)ection pump, or ~a

'IowI pres~

sure safety'nfect'ton pump to the Reactor Coolant Systim.

APPLICABIILITY:- MODES 5 and 6'.

ACTION:

Nth none of the above flow paths OPERABLE', s'uspend all.*operations involving CORE ALTERATIONS or positive reactivi'ty ch'6nges.

.SURVEILLANCE REQUIREMENTS 4.1.2.1,At 'least one o'F 'the above re'quired flow paths shall.be demonstrated OPERABLE at least once per 31, days by verifying that each valve (manual,

'ower-operated, o'r automatic) in the f llaw path t'hat. is not locked, Sea'led, or otherwise secured in position, is,,in its correct position.

PALO VERDE - UNIT'2 3/4 I;6 AMENDMENT NQ. 55

/

RLSS S

LIH T G

0 I I N

FOR OPERATION 3.1.3.4 The individual. full-length (shutdown and regulating)

CEA drop time, from a fully withdrawn position, shall be less than or equal to 4 seconds from when the electrical power,is interrupted to the CEA drive mechanism until the CEA reaches its 90X insertion position with:

a.

T, greater than or equal to 550'F, and b.

All reactor coolant pumps operating.

NODES 1 and 2.,

~ACT I 0 a.

. With the drop time of any full-length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to NODE 1 or 2.

SURVEILLA C RE UIREHENTS 4.1.3.4 The CEA drop.time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a.

For all CEAs following each removal and reinstallation of the:reactor vessel head',

b.

For specifically affected, individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific

CEAs, and c.

At least once per 18 months.

r PALO'ERDE UNIT 2 3/4 1-19 Amendment, No. 55, 100

REACTIVITY CONTROL. SYSTEMS SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdow'n CEAs shall be withdrawn to at'east 144.75. inches,.

APPLICABILITY:

MODES 1 and 2"t.

ACTION:

With a maximum of one shutdown CEA, withdrawn to less than '144.,75 inches~

except for surveillance testing pursuant to Specification 4.1.,3.1.2, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a.

Withdraw the CEA te at 'least 144.75 inches, or b.

Declare the CEA inoperable and'comply with Specification 3. 1.3. 1',

'f ffftt Cf fftVE E

E 4.1.3. 5 Each shutdown CEA shall be determined to be withdrawn to. at least 144.75 inches:

a.

Within 15 Iiinutes prior to withdrawal bf any CEAs in regulating groups duri ng ant approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter except. during time interva'ls

,'hen botlh CEAIC's are inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

See Special Test Exception 3.10.2.

fWith K ff greater than or equal to l.

eff PALO VERDE - UNIT 2

'3/4 1"20 AMENDMENT NO. I 55

POWER DISTRIBUTION LIHITS 3/4. 2. 6 REACTOR COOLANT COLD LEG TEHPERATURE LIHITING CONDITION FOR OPERATION 3.2.6 The reactor coolant cold leg temperature (T ) shall be within the Area of Acceptable Operation shown in Figure 3.2-1.

APPLICABILITY:

HODES 1" and 2"N.

'CTION:

lA'th the reactor coolant cold leg temperature exceeding its limit,, restore the, temperature to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or:be in. HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREHENTS 4.2.6 The reactor coolant cold leg temperature. shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"See Special Test Exception 3.10;4.

SWith K ff greater than or equal to 1 eff PALO VERDE - UNIT 2 3/4 2-7 AHENDHENT NO. 55

575 570 0

565 1-560 555 550 0

545 (30.568) l W T T

570 565

<<1OO,S6O>

5(0 555 550 545-540 J

A I

J 10 20

'30 40 50 60 70 80 90 100

-CORE POWER LEVEL,'YoOF RATED THERMALPOWER (3876 MW)

REACTOR COOLANT COLD LEG TEMPERATURE vs. CORE POWER LEVEL FIGURE 3.2-1 REACTOR.COOLANT COLD LEG TEMPERATURE VS.

CORE,POMER LEVEL PALO VERDE 'NIT 2 3/4 2-8 Amendment No.

55-,M,1OO

EMS BASES and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

Operability of at least two CEA position indicator channels is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits.

The CEA "Full In" and "Full Out" limits provide an additional independent means; for determining the -CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position. indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> s with more frequent verifications required if an automatic monitoring channel is inoperable.

These, verification frequencies are.adequate for assuring that the applicable LCOs are satisfied.

The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses.

measurement with T,~ greater than or equal to 550 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Several design steps were employed to accommodate the possible CEA guide tube wear which could arise from CEA vibrations when ful.ly withdrawn.

Speci,fically, a programmed insertion schedule will be used to cycle the CEAs between the full out position ("FULL OUT" LIMIT) and 3.0 inches inserted over the fuel cycle.

This cycling will distribute the possible guide tube wear over a larger area, thus minimizing any effects.

To accommodate this programmed, insertion schedule, the fully withdrawn position was redefined, in some cases, to be 144.75 inches or, greater.

The establishment of LSSS and LCOs, requires that the expected long-and short-term behavior of the radial peaking factors, be determined.

The long-term behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount of CEA insertion

assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle.

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

The magnitudes of such per.turbations depend upon the expected use of the CEAs during. anticipated power reductions PALO VERDE UNIT 2 B 3/4 1-5 Amendment No.

]Op

REACTIVITY CONTROL,'SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES~iContinue+d and load maneuvering.

Analyses are pei formed'ased on the expected'-

inod@. of operation'of the NSS.'i (Ibase load maneuver'ing, etc:.) and 'from these analyses CEA insertions are de.termined and a-consistent set iof raifial. peaking factors defined.

The Long Term Steady State and Short Term Inseition:L'imits are deter-mined based-upon tlhe assunied mode of operation used i@i the analyse.

and provide a means of preservingl tlie.assumptions on CEA insertions used.

The limits specie fied serve-to, limit -the behavior of the radial peaking factors within the boiinds determined. from analysis.

The actions specifiied serve'o limit the extent of radial xenon redistribution effects to those accommodated:.in the'nalyses.

The',

Long and'Short Term Insertion Limits of, Specifications 3.1.'3.6.and 3.1.3.7're specified for the. plant which. has been designed for primarily,base loaded

'peration but which has the ability to accoagnodate a limiited amount; of load maneuvering.

The Transient Insertion 'Limits. iof Specifications 3.3..3.6 and 3.1.3.'7 the Shutdown CEA Insertion Limit. of.Ipecification 3.1.3.5 ensure that.(1) the minimum SHUTDOW MARGIN is maintained, and (2) the. Pot'ential effects. of..a CEA ejection accident. are limitecl to acceptable levels.

Lon'g-term. operation at the Transient Insertion 'Limiits is niot-permitted'ince such'peration. could have effects on the core power distribution which, could invalidate assumptions used to determine the behavior of the radial peaking factor's..'he

'PVNGS CPC and COLSS systems-are:reslponsible,for the safety and monitoring functions, respect',ively of the reactor core.

COLSS moni'tors the DNB Pow'er Operating Limit (POL) arid various opierating paiamet4rs'd help the oper'ator main-tain plant oper'ation within the limiting conditions for operation.'(LCO).

Dpelrat-

'rig within the LCO guarantees tlhat. in the everi4. of an Anficipated Operatliokal Occurre'nce (A00), the CPCs wi'll prov'ide a relactorl tr'ip'n.time 'ito prevent iin~

acceptable fuel'amage.

The COLSS reserves. the Required OverpowertMargin (ROPM) to account for the Loss. of Flow (LOF) and CEA miisolperation transients.

Shen-the.COLSS,is Out'of, Service (COOS), the monitoring,function.is performed via the CPC ca1culnitibn of DNBR.in conjunction with Technical Slpecification COOS Limit Lines specified in, the -CORE 'OPERATING LIMI'll'S REPORT iihich rest0icts Ithh r'eactor power suff]cihnttly to preser ve the ROPM.

The reduction of'he CEA deviation penalties in accordance with thE.

CEAC (Conti ol Element Assemb'ly Ca'llculator).sensitivity reduction program has'be'en'erfo'rmed.

This task. involved setting many, of the inw'ard single CFA deviation penalty factoi s to 1.'D.

An inward CEA deviation 'event ih effect would not, be iccompanied-by the apiplication of the CEA deviation penalty-in eittier the CPC DNB and.LHR,(Linear Pleat Rate) caflculations for those CEAs with the reduce'd

'penalty factors.

The, protection for an inward CEA deviation event is thus accounted for separately.

PALO VERDE - UNIT,'2

'B 3/4.,1-'6 AMENDMENT NO.

ZS, 55

pe REGS 0

OO IVl O

V/~

O

~O

++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 AR ZONA PUBLIC SERVICE COMPANY ET AL.

DOCK NO.

S N 50-530 LO V RD N

AR GENERAT NG STATION UNIT NO.

3 AM NDMENT TO FACILITY OPERATING LICENSE Amendment No. 80

,License No. NPF-74 1.

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

A.

The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt Ri,ver 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 January 5,

1996, as supplemented by letters dated April 19,

1996, May 1,

1996, and May 10,

1996, complies with the standards and requirements

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

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

D.

.E.

There is reasonable assurance (i) that the activities authorized by this amendment can be c'onducted without endangering the health and safety of the public, and (i'i) that such activiti'es will be conducted in compliance with the Commission's regulations; 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 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.paragraphs 2.C(l) and 2.C(2) of Facility Operating License No.

NPF-74 are hereby amended to read as follows:

Cl 4

im m Power evel 3.

Arizona Public Service Company (APS) is authorized to operate the

.facility at reactor core power levels not in excess of 3876 megawatts thermal (100X power) in accordance with the conditions specified herein and in Attachment 1 to this license.

The items identified in Attachment 1 to this license shall be completed as specified.

Attachment 1 is hereby incorporated into this license.

(2) ec c'c tions and nvironme t 1 Protection Plan The Technical Sperifica~ions contained in Appendix A, as revised through Amendment,No.

80,'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 specifi'c license conditions.

This license amendment is effective as of its date of issuance to be implemented within 30 days of the date of issuance, except for the pressurizer safety valve setpoints change -which is to be implemented prior to startup from Unit 3 refueling outage six.

FOR THE NUCLEAR REGULATORY CONNISSION William T..Russell, Director

'Office of Nuclear, Reactor Regulation Attachments:

1.,

Page 4 of License 2.. Changes to the Technical Specifications Date of Issuance:

May 23, 1996 Page 4 is attached, for convenience, for the composite license to reflect this change.

Please remove page 4 of the existing license and replace with the attached, page.

4

(1)

Arizona Public Service Company (APS) is authorized to operate the facility at. reactor core power levels not in excess of 3876 megawatts thermal (100X power) in accordance with the conditions specified herein and in Attachment 1 to this license.

The items identified in Attachment 1 to this license shall be completed as specified.

Att'achment 1 is hereby incorporated intothis license.

(2) e i

1 e v' e

ot tion Pl n

D.

The Technical Specifications contained in Appendix A, as revised through Amendment 'No.pn 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'lan, except where otherwise stated in specific l.icense conditions.

(3)

An Condition This license is subject to the antitrust conditions delineated in 'Appendix C to this license.

(4).

t r r ecti n14 SR nd R

Any changes in.the initial test program, described in Section 14 of the FSARs (Palo Verde and CESSAR) made in accordance with the provisions of 10 CFR 50.59 shall be reported in accordance with 50.59(b) within one month of such change.

APS has previously been granted an exemption from Paragraph III.D.2(b)(ii) of Appendix J to 10 CFR Part 50.

This exemption. was previously granted in Facility Operating License NPF-65 pursuant to 10 CFR 50.12.

Mith the granting of this exemption, the facility will operate, to the extent authorized herein, in conformity with the application, as

amended, the provisions of the Act, and the rules and regulations of the Commission.

E.

The licensees shall'ully implement and, maintain in effect all provisions of the 'Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Hiscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The Safeguard Contingency Plan is incorporated into the Physical Security Plan.

The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

"Palo Verde Amendment No. Aq

Ik 0

~ ~

0 i[

~ ~

80' C

S C "T'53 M

T G

IC NS 0.:NPF-7 Replace the following pages.of the Appendix A Technical Specifications with the enclosed pages.

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

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

1-5 3/4 1-5 3/4 1-19 3/4 2-8 3/4 4-7'/44-8 B 3/4'.-5 B 3/4 4-12 1-5 3/4 1-'5 3/4 1'-19 3/4 2-,8 3/4 4-7

,3/4 4-8 B 3'/4 1-5 B 3/4 4-12

0

1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the.reactor.

core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

Xf 1.22 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.

1.23 PRESSURE BOUNDARY LEAKA( E sha:1 be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.

0 PC 1.24 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas,

sampling, analyses,

test, and determinations to be made to ensure'hat processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and. other requirements governing the disposal of solid radioactive waste.

1.25 PURGE or, PURGING shall be the controlled, process of discharging air or gas from a confinement to maintain temperature,

pressure, humidity, concentration, or other operating condition,. in such a manner that replacement air or gas is required to purify the confinement.

1.26 RATED THERMAL POWER shall'e a total reactor core 'heat transfer rate to the reactor coolant of 3876 MWt.

R P

N M

1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel.

sensor until electrical power is interrupted to the CEA drive mechanism.

PALO VERDE UNIT 3 1-5 Amendment No. 2-,84,'80

Il 0

~ ~

R C

TI COND TION F RATION

3. 1. 1.4 The Reactor Coolant System lowest operating loop temperature (T<~)

shall be greater than or equal to 545 F.

MGIIEE.I d" 21.

With a. Reactor Coolant System operating loop temperature (T,~) less than 545'F, restore T

~ to within its limit wi.thin 15'inutes or be in HOT STANDBY within the next 5 minutes.

SURVEI ANCE RE UIREMENTS

4. 1. 1.4 The Reactor Coolant System temperature

(~T,~).shall be determined to be greater than or equal to 545 F:

a.

Within 15 minutes prior to achieving reactor criticality, and b.

At least once per 30 minutes when the reactor is cr'itical and the Reactor Coolant System T,~ is less than 550'F.

PWith K,<< greater than or equal to 1.0.

PALO VERDE UNIT 3

'3/4 1-5 Amendment No. 27-,42-,49,.nn

3/4.1.2.

BO iON S'(STEMS FLOW PATMS " SMUTDOWN LIMITING.COINDITION POR OPERATION 3.1.2.1 As a mini!mum, one of the fdlldvihg bor'on 'infection flow paths shall be OPERABLE,':

a.

If only the spent fuel pool in!ipeicification 3.1.2.5a.

is OPERABLE, a flow path

~From the spent'uel pool via a.gravity feed connection and a charging Ipaep t~ the Reactor Coolant System.

b.

If only the refueling water tank in SIpecification 3.1.2.5b'., is OPER%LE, a flow path from the mfueling water tank via either a charging pump, i high pressure safety infection pmap, or a 'low pres-sure safety injection pump to th'e Reactor Coolant System~

APPLICABI:LITY:

MODES 5 and 6.

ACTION:

With none of the above flow paths OPIERABLE, suspend all operations involving CORE ALTERA1'IONS or positive reactivtity changes.

SURVEILLANCE REtLUIREMENTS 4.1.2.1 At least one of 'the above required flow paths sha'll be demonstrated OPERABLE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not:locked,

sealed, or otherwise securedl in position, is in its correct position.

PALO VERDE - UNIT 3 3/4 1-6

QKhlDMENT, NO. 42

L T

G CO D

0 0

0 ERAT ON

3. 1.3.4 The individual full'-length (shutdown and regulating)

CEA drop time, from a fully withdrawn position, shall be less than or equal to 4 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90X insertion position with:

a.

T,,

greater

.than or equal to 550 F, and b.

All reactor coolant pumps operating.

NODES 1 and 2.

KITH a ~

Mith the drop time of any ful,l-length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to NODE 1 or 2.

SURVEILLANCE R IREHENTS'.

1.3.4 The CEA drop time of full'-length CEAs shall be demonstrated through measurement prior to reactor criticality:

a.

For. all CEAs following each removal and reinstallation of the reactor vessel

head, b.

For specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and c.

At least once per 18 months.

PALO VERDE UNIT 3 3/4 1-19 Amendment No. 48,80

SHUTDOWN CEA INSERTION LIMI1l LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to -at least 144.75 inches.

APPLICABILITY:

MODES 1 and 2"¹.

ACTION:

With a maximum of.one shutdown CEA withdraw'n to less than 144.75 inches,,

except for surveiIllance testing pursuant to'~'iec'ification 4.1.3.1.2, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either:

a.

Withdraw the CEA to at least 144.75 inches, or b.

Declare the CEA inoperable and-comply lith Specification 3.1.3.1.

SURVEILLANCE RE UIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to:be withdrawn 'to.at lelastt 144. 75 inches:

a.

Within 15 minut!es prior to withdrawal of any CEAs in regulating groups during ain aipproach to reactor, criticalityand b.

At.least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter except during time intervals when both CEAC'.s aire inoperable, then verify the individual CEA positions at least once. per 4

-hou'rs.'See Special Test Exception 3.10.,2.

¹With K ff greater than or equal to l.

eff PALO VERDE - UNIT 3 3/4 1-20 ANENCNENI NO. ZN, AZ

POWER DISTRIBUTION LIMITS 3/4.2. 6 REACTOR COOLANT COLD LEG TEMPERATURE LIMITING CONDITION FOR OPERATION 3.2. 6 The reactor coolant cold leg temperature (T ) shall be within the Area of Acceptable Operation shorn in Ffgure 3.2-1.

APPLICABILITY:

NODES 1* and 2"f.

ACTION:

With the 'reactor coolant cold leg temperature exceeding its lfaft, restore the temperature to v$thfn its lfmft within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be fn HOT STANDBY &thin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS

4. 2. 6 The reactor coolant cold leg temperature shall be determined to be

~ithin its lfsit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

"See Special Test Exceptfon 3.10.4.

NMfth K ff greater than or equal to 1 eff PALO VERDE - UNIT 3 3/4 2-7 AMENDMENT NO.. 42

575

.570 O

565 I-560 555 55o 0O 545 570-56,'5 (I00,560)

'60 55,'5 550 545 540 I

I 10 20 30 40 50 60 70 80 90 100 CORE POWER LEVEL;% OF RATED THERMAL.POWER (3876 MW)

REACTOR COOLAI'VTCOLD LEG TEI41PERATURE v's. CORE POWER LEVEL F'IGURE 3.2-1 REACTOR COOLANT COLD LE'G TEMPERATURE VS.

CORE POWER LEVEL

.PALO VERDE UNIT 3 3/4 2,'8 Amendment No. iQ-,498O

C 0 S

TI G

C N

TION FOR OPERATION 3.4.2. 1 A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2475 psia +3, -1X*.

MODE 4.

'a ~

b.

Mith no pressurizer code safety valve OPERABLE, immediately suspend all operations

nvolving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.

The provisions of Specification 3.0.4 may be suspended.

for, up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for entering into and during operation in MODE 4 for purposes of setting the pressurizer code safety valves under ambient (HOT) conditions provided a preliminary cold setting was made prior to heatup.

SU VE LLA E UI EMENTS 4.4.2.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.

• The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

PALO VERDE UNIT 3 3/4 4-7 Amendment No. 47-, Bl)

SQJ~

~NCE Tl II~IEELEE%J)ITI N

3.4.2.2 All pressurizer code safety valves shall be OPi;RABLE with a lift setting of 2475 p.sia +3, -IX*.

~C~O:

Mi'th one pressurizer code safety valve inoperable, either. restore the inoperable valve to OPI=RABLE status within ~15~mi~nutes or be in at least, MOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in MOl SIHUTDOMN iwithiin the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with the =shutdown cool'ing,system suction line relief valves aligned to provide overpressure protection for the Reactor Coolant

~System.

~RTEILL NCE RE El~@IT~I 4.4.2.2 No additional Surveillance Requirements other, than those required by Specification 4.05.

• The lift setting pressure

.'hall'orrespond to a'mbient conditions of the val,'ve,

,at nominal operating temperature and pressure.

PALO VERDE

. UNIT 3 3/4 4-8 Amendment No. 47-,80-

T 0 BSS Cont and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude. such adverse conditions from developing.

Operability of at least two CEA position indicator channels is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits.

The CEA "Full In" and "Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.

CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

The maximum CEA drop time restriction is consistent with the assumed CEA drop time used in the safety analyses.

Measurement with T,~ greater than or equal to 550 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Several design steps were employed to accommodate the possible CEA guide tube wear which could arise from CEA vibrations when fully withdrawn.

Specifically, a programmed insertion schedule will be used to cycle the CEAs between the full out position

(".FULL OUT" LIMIT) and 3.0 inches inserted over the fuel cycle.

This cycling will distribute the possible guide tube wear over a larger area, thus minimizing any effects.

To accommodate this programed insertion schedule, the fully withdrawn position was redefined, in some cases, to be 144.75 inches or greater.

The establishment of LSSS and LCOs requires that the expected long-and short-term behavior of the radial, peaking factors be determined.

The long-term behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount of CEA insertion

assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle.

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

The magnitudes of such perturbations depend upon the expected use of the,CEAs during anticipated power reductions PALO VERDE UNIT 3

'B 3/4 1-5 Amendment No.8n'

0 REACTIVITY CONTROL SYSTEM!i BASES MOVABLE CONTROL ASSEMBLIES (Continued) and load maneuvering,.

Analyses are'erformed based on the expected module of operation of the NISSS (ba. e load maneuvering, et<.) ahd from these analyses CEA insertions are determined and a consistient set of r'adial peaking factors, defined.

The Long Term Steady State and Shirt, Term Insertion"Limits are deter-mined based upon the assumed mode of operation used in the analyses.

and prtovide a means, of preserving the assumptions on CEA i'nskrtiohs used.

The limits spieci-fied serve to limit 'the.behavior of the radial, peaking factors within the boundls determined from analysis.

The actions specified-serve to limit the extient of radial xenon redistribution effects to thos~ accommodated i'n the analyses.

The Long and'Short. Term Insertion iimit. of Specif'ications 3.1.3.6 and 3.1.3.7 are specified for the plant which has been designed for primarily base

loaded, operation but whic:h has the ab.ility to accoeeodate a 'llimited amount of load maneuvering.,

The Transient Insertion Limits of Specifications, 3.1.3.'6 arid'.1.3.7 and the Shutdown CEA Insertion Limits of Specification 3.1. 3.5 ensure that (1) the i

minimum SHUTDOWN NiARGINI is maintained, and (2) the potential. effects ofi a iCEA'jection accident are limited to acceptable leve'is..

Long-.term operation at the Transient Inserticin ILimits is not, permitted since such operation could 'Ihave effects on the core power distribution which-cou')d invalidate assumptionsused to determine the behavior of the radial'eaking factors.

The PVNGS CPC and COLSS systems are responsible for the safety and, mbnitoring functions, respectively, of thee re~ctor core.

COLSS monitors the DNB Power Operating Liait (POL) and various operating paratoeters to help the operator main-

.tain plant operation within thee limiting conditions for operation (LCO)'.

IOpierat-ing within the LC() guarantees that in the'vent of an Anticipated Operational Occurrence (AOO), the CPCs will provide a rleator trip in time to prevent;un; acceptable fuel damage.

The COLSS r'e. erye.

the Required Overpower Margin (ROPM) to accountj for Loss of Flow (LOF) and CEA misoperation trans'ients.

lichen the COLSS is',Out qf Service (COOS), tice monitor'ing function is iperfoirmed via tlhe CPC calculation of" DNBR in conjunction with Technical Specificatio'n COOS. Limit Linis specified in the CORE OPERATING LIMITS REPORT which restricts the reactor power suff,iciently to,preserve the ROPM.

The reduction of 'the CEA deviation penalties in accor'dance wi'th the CEAC (Control Element Assembly Calculator),sensfitivity reduction program has been performed.

This task involvedl setting, many, of the inward single CEA deviation penalty factors to 1.0.

An inward CEA deviation event in effect would not,be accompanied by the application o'f the CEA deviation penalty in either the CPC DNB and LHR (Linear Heat Rate) calculations for~those CEAs with the reduced penalty factors.

The protectilon for an inwat d CEA deviation event is thus accounted for separately.

PALO VERDE - UNIT 3 B 314;1-6

<<AMENDMENTINO XS.~ 42

B S

I S

onti ed A typical LTOP system includes pressure relieving devices and a number of administrative and operational controls.

Each of the Palo Verde Units has a similar LTOP system that includes two Shutdown Cooling System suction line relief valves for transient mitigation.

Each relief valve has an opening setpoint of 467 psig which, in combination with certain other limiting conditions for operation contained in Technical Specifications, comprises the LTOP system.

Previously, the LTOP enable temperature during heatup and cooldown have been determined at the intersections between a horizontal line corresponding to the safety valve setpoint (2475 psia) and the most limiting P-T l.imit curves for heatup and cooldown, respectively.

Note that the enable temperature generally identifies the upper temperature limit below which. the LTOP system has to be operable.

In this. analysis, the LTOP enable temperatures were determined in accordance with a definition contained in the latest revision of the Standard Review Plan 5.2.2.

According to SRP 5.2.2 the LTOP enable temperature is "the water temperature corresponding to a metal temperature of at least RT~~ + 90'F at the beltline location (I/4T or 3/4T) that is control-ling in the Appendix G limit calculations."

The heatup and cooldown rate limitations assure the limits of Appendix G to 10 CFR 50 will not be exceeded with overpressure protection provided by the primary safety valves.

The various categories of load cycles used for design purposes are provided in Chapters 3

and 5 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so as not to exceed the limit lines of Figures 3.4-2a and 3.4-2b.

This ensures that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

PALO VERDE UNIT 3 B 3/4'-12 Amendment No. 84, 80.

~I