Text

Amend 24 to License NPF-41,revising Tech Specs to Incorporate Changes in Support of Cycle 2 Operation
	ML17300B099
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Site:	Palo Verde
Issue date:	10/21/1987
From:	Knighton G; Office of Nuclear Reactor Regulation
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UNITED STATES NUCLEAR REGULATORY COPIMISSION WASHINGTON, D. C. 20555 ARIZONA PUBLIC SERVICE COMPANY ET AL.

DOCKET NO.

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

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

24 License No. NPF-41 1.

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

A.

The application for amendment, dated June 29, 1987, as supplemented by letters dated June 29, July 13, August 20 (two letters),

Septem-ber 4 and October 1, 1987, by the Arizona Public Service Company (APS) 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 (licensees),

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 Act, and the regulations of the Commission; C.

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

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

8710270363 871021 PDR ADOCK 0500052S',

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2.

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

(2)

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

24, and the Environmental Protection Plan contaired 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.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION George

. Knighton, irector Projec Directorate V

Division of Reactor Projects - III, IV, V and Special Projects Office.of Nuclear Reactor Regulation

Enclosure:

Changes to the Technical Specifications Date of Issuance:

October 21, 1987

October 21, 1987 ENCLOSURE TO LICENSE AMENDMENT AMENDMENT NO. 24 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.

Also to be replaced are the following overleaf pages to the amended pages.

~dd A

IV XIX XX 2-1 2-3 2-5 2-1 2-2 2~3 2-6 3/4 1-2a 3/4 1-5 3/4 1-21 3/4 1-22 3/4 1-23 3/4 1-24 3/4 1-25 3/4 1-31 3/4 1-32 3/4 1-33 3/4 1-34 3/4 2-1 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-7a 3/4 2-8 3/4 2-12 3/4 3-7 3/4 3-8 3/4 3-9 3/4 3-10 3/4 3-11 3/4 3-12 3/4 3-13 3/4 3-26 Overleaf Paqes 2-2 2-4 2-6 B 2-4 3/4 1-6 3/4 1-26 3/4 2-2 3/4 2-11 3/4 3-14 3/4 3-25

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ENCLOSURE TO LICENSE AHENDYiENT CONTINUATION Amendment Pa es B 3/4 1-6 B 3/4 1-7 B 3/4 2<<1 B 3/4 2-3 8 3/4 2-4 B 3/4 3-1 B 3/4 3-2 5-5 Ove~leaf Pages

. B 3/4 1-5 B 3/4 2-2 5-6

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

2. 1. 1 REACTOR CORE

2. 1. 1. 1 DNBR.

2. 1. 1. 2 PEAK LINEAR HEAT RATE...

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE 2.2 LIHITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETPOINTS...................

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

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4. 0 APPLICABILITY..

3/4. 1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 BORATION CONTROL PAGE 3/4 0-1 SHUTDOWN MARGIN - ALL CEAs FULLY INSERTED............

3/4 1-1 SHUTDOWN MARGIN KN 1 ANY CEA WITHDRAWN MODERATOR TEMPERATURE COEFFICIENT...........

MINIMUMTEMPERATURE FOR CRITICALITY.

3/4. 1. 2 BORATION SYSTEMS 3/4 1-2 3/4 1"4 3/4 1-6 FLOW PATHS " SHUTDOWN.

FLOW PATHS - OPERATING..........

CHARGING PUMPS " SHUTDOWN.....................'........

CHARGING PUMPS - OPERATING.

BORATED WATER SOURCES SHUTDOWN.

BORATED WATER SOURCES " OPERATING.;...................

BORON DILUTION ALARMS 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-10 3/4 1-11 3/4 1-13 3/4 1-14 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION..................

POSIITION INDICATOR CHANNELS " OPERATING..............

POSITION INDICATOR CHANNELS " SHUTDOWN................

CEA DROP TIME...

SHUTDOWN CEA INSERTION LIMIT..........................

REGULATING CEA INSERTION LIMITS PART LENGTH CEA INSERTION LIMITS......................

I 3/4 1-21 3/4 1-25 3/4 1-26 3/4 1"27 3/4 1-28 3/4 1-29 3/4 1-33 PALO VERDE - UNIT 1 IV AMENDMENT NO.

24

LIST OF FIGURES INDEX PAGE

3. 1-1A

3. 1-1

3. 1-2

3. 1-2A

3. 1-3

3. 1-4

3. 1-5

3. 2-1

3. 2"2 3.2"2A 3.2 3

3.4"1

3. 4"2

4. 7-1 SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............

3/4 1-2a ALLOWABLE MTC MODES 1 AND 2..

MINIMUM BORATED WATER VOLUMES CORE POWER LIMIT AFTER CEA DEVIATION.

CEA INSERTION LIMITS VS THERMAL POWER (COLSS IN SERVICE).

3/4 1-.5 3/4 1-12 3/4 1-24 3/4 1-31 CEA INSERTION LIMITS VS THERMAL POWER (COLSS OUT OF SERVICE).

3/4 1-32 PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER.......

3/4 1-34 COLSS DNBR POWER OPERATING LIMIT ALLOWANCE FOR BOTH CEACs INOPERABLE.....

3/4 2"6 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE).....

3/4 2-7 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE)...

3/4 2-7a REACTOR COOLANT COLD LEG TEMPERATURE'S CORE POWER LEVE Le

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3/4 2-10 DOSE E(UIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY

> 1.0 pCi/GRAN DOSE E(UIVALENT I"131...................

3/4 4"28 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER OPERATION..............

3/4 4"30 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..............

3/4 7-26 B 3/4.4-1 NIL-DUCTILITYTRANSITION TEMPERATURE INCREASE AS FUNCTION OF FAST (E > 1 MeV) NEUTRON FLUENCE B 3/4 4-10

5. 1-1

5. 1-2

5. 1-3

6. 2-1

6. 2"2 SITE AND EXCLUSION BOUNDARIES...............

5-2 LOW POPULATION ZONE

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5-3 GASEOUS RELEASE POINTS.................................

5-4 OFFSITE ORGANIZATION..................................

6-3 ONSITE UNIT ORGANIZATION...............................

6"4 PALO VERDE - UNIT 1 XIX AMENDMENT NO.

24

LIST OF TABLES INDEX 1.2

2. 2-1 FREQUENCY NOTATION OPERATIONAL MODES REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS PAGE

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1 8 1-9 2-3

3. 1-1

3. 1-2

3. 1-3

3. 1-4

3. 1"5

3. 3-1

3. 3-2

4. 3-1

3. 3-3

3. 3-4 3.3-5 4.3-2 3.3-6 4.3-3

3. 3-7

4. 3-4 3.3-8 4.3-5 3.3-9 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGIN PUMPS AND PLANT OPERATIONAL MODES....................

FOR K ff > 0.98.

FOR 0.98 K ff > 0.97 FOR 0.97 K ff ) 0.96.

FOR 0.96 K

> 0.95 FOR K ff < 0.95..

REACTOR PROTECTIVE INSTRUMENTATION.............~.....

REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES....

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES..........................

ENGINEERED SAFETY FEATURES

RESPONSE

TIMES ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS RADIATION MONITORING INSTRUMENTATION~

RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................

SEISMIC MONITORING INSTRUMENTATION.

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS METEOROLOGICAL MONITORING INSTRUMENTATION.

METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.

REMOTE SHUTDOWN INSTRUMENTATION, DISCONNECT..........

SWITCHES AND CONTROL CIRCUITS 3/4 1"16 3/4 1-17 3/4 1-18 3/4 1-19 3/4 1-20 3/4 3-3 3/4 3-11 3/4 3-14 3/4 3"18 3/4 3"25 3/4 3-28 3/4 3-31 3/4 3-38 3/4 3-40 3/4 3-43 3/4 3-44 3/4 3"46 3/4 3-47 3/4 3-49 PALO VERDE " UNIT 1 XX AMENDMENT NO. 24

l, 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTE SETTINGS

- 2. 1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR

2. 1. 1.1 The calculated DNBR of the reactor core shall be maintained greater than or equal to 1.24.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the calculated DNBR of the reactor has decreased to less than 1.24, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specifi-cation 6.7.1.

PEAK LINEAR HEAT RATE 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be maintained less than or equal to 21 kW/ft.

APPLICABILITY:

MODES 1 and 2.

ACTION'henever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE

2. 1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY:

MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1

hour, and comply with the requirements of Specification 6.7. 1.

MODES 3, 4, and 5:

'henever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1.

PALO VERDE - UNIT 1 2-1 AMENDMENT NO. 24

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2. 1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1

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ACTION:

With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.

1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

PALO VERDE - UNIT 1 2-2

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

TRIP GENERATION A.

Process 1.

Pressurizer Pressure

- High 2.

Pressurizer Pressure

- Low 3.

Steam Generator Level - Low 4.

Steam Generator Level - High 5.

Steam Generator Pressure

- Low 6.

Containment Pressure

- High 7.

Reactor Coolant Flow - Low a.

Rate b.

Fl oor c.

Band 8.

Local Power Density High 9.

DNBR " Low B.

Excore Neutron Flux 1.

Variable Overpower Trip a.

Rate b.

Cei 1 ing c.

Band TRIP SETPOINT

< 2383 psia

> 1837 psia (2)

> 44.2X (4)

< 91.0X (9)

> 919 psia (3)

< 3.0 psig

< 0.115 psi/sec (6)(7)

> 11.9 psid (6)(7)

< 10.0 psid (6)(7)

< 21.0 kW/ft (5)

> 1.24 (5)

< 10.6X/min of RATED THERMAL POWER (8)

< 110.0X of RATED THERMAL POWER (8)

< 9.8X of RATED THERMAL POWER (8)

ALLOWABLE VALUES

< 2388 psia

> 1822 psia (2)

> 43.7X (4)

< 91.5X (9)

> 912 psia (3)

< 3.2 psig

< 0.118 psi/sec (6)(7)

> 11.7 psid(6)(7)

< 10.2 psid (6)(7)

< 21.0 kW/ft (5)

> 1.24 (5)

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

< 111.0X of RATED THERMAL POWER (8)

< 10.0X of RATED THERMAL POWER (8)

TABLE-2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOIHT LIMITS ID m

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FUNCTIONAL UNIT 2.

Logarithmic Power Level - High (1) a.

Startup and Operating b.

Shutdown C.

Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators D.

Supplementary Protection System Pressurizer Pressure - High II.

RPS LOGIC A.

Matrix Logic B.

Initiation Logic III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manukl 'Trip TRIP SETPOINT

< 0.010X of RATED THERMAL POWER

< 0.010K of RATED THERMAL'POWER Hot Applicable Hot Applicable

< 2409 psia Hot Applicable Hot Applicable Not Applicable Not Applicable ALLOWABLE VALUES

< 0.011K of RATED THERMAL POWER

< 0.011K of RATED THERMAL POWER Hot Applicable Hot Applicable

< 2414 psia Hot Applicable Not Applicable Not Applicable Not Applicable ED 44

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1)

Trip may be manually bypassed above 10-4X of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10"4X of RATED THERMAL POWER.,

(2)

In MODES 3-4, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached.

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

(3)

In MODES 3-4, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4)

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

(5)

As stored within the Core Protection Calculator (CPC).

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

Trip may be manually bypassed below 10-4X of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 10-~X of RATED THERMAL POWER.

PALO VERDE - UNIT 1 2-5 AMENDMENT NO.

24

TABLE 2.'2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (Continued)

(6)

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

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

FLOOR is the minimum value of the trip setpoint.

SANO is the amount by which the trip setpoint is below the input signal

~un ess limited by Rate or Floor.

Setpoints are based on steam generator differential pressure.

(7)

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

(8)

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

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

CEILING is the maximum value of the trip setpoint.

BAND >s the amount by which the trip setpoint is above the steady state input signal unless limited by the rat% or the ceiling.

(9)

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

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.---; eeoMENT Noe 19

2.1 and 2.2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES

2. 1. 1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21 kW/ft which will not cause fuel centerline melting in any fuel rod.

First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.

The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).

At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.

Correlations predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DNB.

The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.24 based upon a statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.

The DHBR limit of 1.24 includes a rod bow compensation of 1.75K on DNBR.

Second, operation with a peak linear heat rate below that which would cause fuel center line melting maintains fuel rod and cladding integrity.

Above this peak linear heat rate level (i.e., with some melting in the center),

fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods.

Volume changes which accompany the solid to liquid phase change are significant and require accommodation.

Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting.

Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit.

To account for fuel rod dynamics (lags), the directly indicated linear heat rate is dynamically adjusted by the CPC program.

PALO VERDE - UNIT 1 B 2-1 AMENDMENT NO.

24

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BASES Limiting Safety System Settings for the Low DNBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and kW/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.

2. 1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The Reactor Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of llOX (2750 psia) of

-design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.

The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of 1.24 and 21 kW/ft, respectively.

Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment.

The Allowable Values for these trips are therefore the same as the Trip Setpoints.

To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DNBR - Low and Local Power Density High trips include the measurement, calculational and processor uncertainties and dynamic allowances as defined in the latest applicable revision of CEN-305-P, "Functional Design Requirements for a Core Protection Calculator,"

and CEN-304-P, "Functional Design Requirements for a Control Element Assembly Calculator."

PALO VERDE " UNIT 1 B 2"2 AMENDMENT NO. 24

BASES REACTOR TRIP SETPOINTS (Continued)

The methodology for the calculation of the PVNGS trip setpoint values, plant protection system, is discussed in the CE Document No. CEN-286(V),

Rev.

2, dated August 29, 1986.

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

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

This trip function

-'will trip the reactor when the indicated neutron flux power exceeds either a

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

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

These trip setpoints are provided in Table 2.2-1.

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

A reactor trip is initiated by the Logarithmic Power Level High trip unless this trip is manually bypassed by the operator.

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

Pressurizer Pressure

- Hi h The Pressurizer Pressure

- High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of loss of load without reactor trip.

This trip s setpoint is below the nominal liftsetting of the pressurize~

safety valves and its operation minimizes the undesirable opera-tion of the pressurizer safety valves.

Pressurizer Pressure

- Low The Pressurizer Pressure - Low trip is provided to trip the reactor and

- to assist the Engineered Safety Features System 'in the event of a decrease in Reactor Coolant System inventory and in the event of an increase in heat PALO VERDE - UNIT 1 B 2-3

, AMENDMENT NO. 24

SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES Pressurizer Pressure Low (Continued) removal by the secondary system.

During normal operation, this trip's set-point may be manually.decreased, to a minimum value of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached.

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

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

Containment Pressure

- Hi h

The Containment Pressure

- High trip provides assurance that a reactor trip is initiated in the event of containment building pressurization due to a pipe break inside the containment building.

The setpoint for this trip is identical to the safety injection setpoint.

Steam Generator Pressure Low The Steam Generator Pressure - Low trip provides protection in the event of an increase in heat removal by the secondary system and subsequent cooldown of the reactor coolant.

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

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

Steam Generator Level -

Low The Steam Generator Level - Low trip provides protection against a loss of feedwater flow incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to a decrease in heat removal by the secondary system.

This specified setpoint provides allowance that there will be sufficient water inventory in the steam generator at the time of the trip to provide a margin of at least 10 minutes before auxiliary feedwater is required to prevent degraded core cooling.

Local Power Densit

- Hi h The Local Power Density - High trip is provided to prevent the linear heat rate (kW/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any design bases anticipated operational occur-rence.

The local power density is calculated in the reactor protective system utilizing the following information:

PALO VERDE " UNIT 1 B 2-4

BASES Local Power Densit

- Hi h (Continued) a.

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

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

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

The loca')

power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines.

These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result

.in a violation of the Peak Linear Heat Rate Safety Limit.

CPC uncertainties related to peak LPD are the same types used for DNBR calculation.

Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density),

sensor time delays, and protection system equipment time delays.

DNBR - Low The DNBR -

Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event

'f design bases anticipated operational occur rences.

The DNBR - Low trip incorporates a low pressurizer pressure floor of 1860 psia.

At this pressure a

DNBR - Low trip will automatically occur.

The DNBR is calculated in the CPC utilizing the following information:

a.

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

Reactor Coolant System pressure from pressurizer pressure measurement; C.

Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements; d.

e.

Radial peaking factors from the position measurement for the CEAs; Reactor coolant mass flow rate from reactor coolant pump speed; Core inlet temperature from reactor coolant cold leg temperature measurements.

PALO VERDE - UNIT 1 B 2-5 AMENDMENT NO.

24

SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES DNBR Low (Continued)

The DNBR, the trip variable, calculated by the CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is initiated prior to violation of fuel design limits.

These uncertainties and dynamic compensa-tion routines ensure that a reactor trip occurs when the calculated core DNBR is sufficiently greater than 1.24 such that the decrease in calculated core DNBR after the trip will not result in a violation of the DNBR Safety Limit.

CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties.

Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.

The DNBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.

a@

b.

C.

d.

e.f.

g.

h.

Parameter RCS Cold Leg Temperature-Low RCS Cold Leg Temperature-High Axial Shape Index-Positive Axial Shape Index-Negative Pressurizer Pressure-Low Pressurizer Pressure-High Integrated Radial Peaking Factor-Low Integrated Radial Peaking Factor-High guality Margin-Low Limitin Value

> 470 F

< 610 F

Not more positive. than

+ 0.5 Not more negative than - 0.5

>.1860 psia

< 2388 psia

> 1.28

< 4.28

> 0 Steam Generator Level Hi h The Steam Generator Level' High trip is provided to protect the turbine from excessive moisture carry over.

Since the turbine is automatically tripped when the reactor is tripped, this trip provides a reliable means for providing protection to the turbine from excesssive moisture carryover.

This trip's setpoint does not correspond to a safety limit, and provides protection in the event of excess feedwater flow.

The setpoint is identical to the main steam isolation setpoint.

Its functional capability at the specified trip setting enhances the overall reliability of the reactor protection system.

PALO VERDE - UNIT 1 B 2-6 AMENDMENT NO.

24

I II (509,6.5)-

III

~

p I

IIII I

III I

II I

I II I

P I

I I

5'EGION OF ACCEPTABLE--'--------

OPERATION I

I II II III I

II z

CD CL I

I I

(350 3 5)---'-------'

I REGION OF II I

I'NACCEPTABLE OPERATION I

2' IJ

~

I I

0' IOO 200 300 100 500 600 24 AMENDMENT NO.

3/4 1-2a PALO VERDE " UNIT 1 COLD LEG TEMPERATURE ('F)

FIGURE 3.I-IA SHUTDONfN MARG'lN VERSUS COLD LEG TEMPERATURE

l

~0/,0.5)

FIGURE 3.1-1 ALLOWABLE MTC MODES 1 AND 2 PALO VERDE UNIT 1 CYCLE 2 0.5 0

IIOO/.,O.O)

I U

LLI D

LLI CL cf 0

LIJ CL l

-0.5

-l.5 At,,LOWABLE MTC C)

I 0

C)

-2.5 IO/.-2.8)

(IOO/,-3.5)

-3.5 0

20 40 60 CORE POWER LEVEL,/ OF RATED THERMAL POWFR 80 IOO

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION

3. l. 1.4 The 'Reactor Coolant System lowest operating loop temperature (T

ld) shall be greater than or equal to 552 F.

APPLICABILITY:

MODES 1 and 2¹*.

ACTION:

With a Reactor Coolant System operating loop temperature (T

ld) less than cold 552 F, restore T

ld to within its limit within 15 minutes or be in HOT cold STANDBY within the next 15 minutes.

SURVEILLANCE RE UIREMENTS

4. 1. 1.4 The Reactor Coolant System temperature (T

ld) shall be determined to be greater than or equal to 552'F:

cold 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

ld is less than 557'F.

cold

¹With K ff greater than or equal to 1.0.

eff "See Special Test Exception 3. 10.5.

PALO VERDE - UNIT 1 3/4 1-6

'REACTIVITY CONTROL SYSTEMS 3/4. 1. 3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION

3. 1.3.1 All full-length (shutdown and regulating)

CEAs, and all part-length CEAs which are inserted in the core, shall be OPERABLE with each CEA of a given group positioned within 6.6 inches (indicated position) of all other CEAs in its group.

APPLICABILITY:

MODES 1* and 2".

ACTION:

a ~

b.

C.

With one or more full-length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With more than one full-length or part-length CEA inoperable or misaligned from any other CEA in its group by more than 19 inches (indicated position),

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

With one or more full-length or part-length CEAs misaligned from any other CEAs in its group by more than 6.6 inches, operation in MODES 1 and 2 may continue, provided that core power is reduced in accordance with Figure 3. 1-2A and that within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the misaligned CEA(s) is either:

1.

Restored to OPERABLE status within its above specified alignment requirements, or 2.

Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

After declaring the CEA(s) inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specifications 3.1. 3. 6 and 3.1. 3. 7 provided:

a)

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the remainder of the CEAs in the group with the inoperable CEA(s) shall be aligned to within 6.6 inches of the inoperable CEA(s) while maintaining the allowable CEA sequence and insertion limits shown on Figures 3.1-3 and 3.1-4; the THERMAL POWER level shall be restricted pur-suant to Specifications 3.1.3.6 and 3. 1.3.7 during subse-quent operation.

"See Special Test Exceptions

3. 10.2 and 3.10.4.

PALO VERDE - UNIT 1 3/4 1-21 AMENDMENT NO. 24

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) b)

The SHUTDOWN MARGIN requirement of Specification

3. l. 1. 1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 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 .

d.

e.

With one full-length CEA inoperable due to causes other than addressed by ACTION a.,

above, but within its above specified align-ment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification

3. 1.3.6.

With one part-length CEA inoperable and inserted in the core, operation may continue provided the alignment of the inoperable part length CEA is maintained within 6.6 inches (indicated position) of all other part-length CEAs in its group and the CEA is maintained pursuant to the requirements of Specification

3. 1. 3. 7.

SURVEILLANCE RE UIREMENTS 4.1.3.1.1 The position of each full-length and part-length CEA shall be determined to be within 6.6 inches (indicated position).of all other CEAs in its group at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4. 1.3. 1.2 Each full-length CEA not fully inserted and each part-length CEA which is inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.

,PALO VERDE " UNIT 1 3/4 1-22 AMENDMENT NO. 24

THIS PAGE INTENTIALLY LEFT BLANK PALO VERDE - UNIT 1 3/4 X-23 AMENDMENT NO.

24

z0l-o+

~ llJ mo 30 O

~ Q.

20 O~

0Ol 10 D I-ac 0

CC ii 20

~

~ ~ ~

r i

=

~

I (60 MIN, 20%)

0 10 20 z

TIME AFTER 30 40 50 60 DEVIATION,MINUTES WHEN CORE POWER IS REDUCED TO 55% OF RATED THERMAL POWER PER THIS LIMITCURQE, FURTHER REDUCTION IS NOT REQUIRED PALO VERDE UNIT 1 FIGURE 3.1-2A CORE POWER LIMIT AFTER CEA DEVIATION*

3/4 1-24 AMENDMENT NO.

24

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS " OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA:

a.

CEA Reed Switch Position Transmitter (RSPT 1) with the capability of determining the absolute CEA positions within 5.2 inches, b.

CEA Reed Switch Position Transmitter (RSPT 2) with the capability of determining the absolute CEA positions within 5.2 inches, and c.

The CEA pulse counting position indicator channel.

APPLICABILITY:

MODES 1 and 2.

ACTION:

Mith a maximum of one CEA per CEA group having only one of the above required CEA position indicator channels OPERABLE, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

Restore the inoperable position indicator channel to'PERABLE

status, or b.

C.

Be in at least HOT STANDBY, or Position the CEA group(s) with the inoperable position indicator(s) at its fully withdrawn position while maintaining the requirements of Specifications

3. 1. 3. 1, 3. 1. 3. 6 and 3. 1. 3. 7.

Oper ation may then continue provided the CEA group(s) with the inoperable position indicator(s) is maintained fully withdrawn, except during surveill-ance testing pursuant to the requirements of Specification 4. 1.3. 1.2, and each CEA in the group(s) is verified fully withdrawn at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter by its "Full Out" limit."

SURVEILLANCE RE UIREMENTS

4. l. 3. 2 Each of the above required position indicator channels shall be determined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5.2 inches of each other at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"CEAs are fully withdrawn (Full Out) when withdrawn to at least 144.75 inches.

PALO VERDE - UNIT 1 3/4 1-25 AMENDMENT NO.

24

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS -

SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.3.3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERABLE for each

shutdown, regulating or part-length CEA not fully inserted.

APPLICABILITY:

MODES 3*, 4*, and 5*.

ACTION:

With less than the above required position indicator channel(s)

OPERABLE, immediately open the reactor trip breakers.

SURVEILLANCE RE UIREMENTS

4. 1.3.3 The above required CEA Reed Switch Position Transmitter indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.

With the reactor trip breakers in the closed position.

PALO VERDE - UNIT 1 3/4 1-26

nO It/I C/l Dl n

Pl 0.90 0.80 o

0.70 0.60 0.50 0,40 0.30 0.20 O.IO 0.00 C/l C7 W D mM

~ CJl m<

~ D D Co Igo

-4 'D C)

~ D 5 M~

mm D ~ gl

~mo D

TRANSIENT INSERTION LIMIT o

150 l20 90 60 30 0

l50 l20 90 60 30 0

2 150 I20 90 60 30 0

I50 I20 90 60 30 0

CEA Y/ITHDRAYIAL-INCHES

0.90 ill Pl KI(

fTl C)

Pl 0.80 nxr g

070 Kl C) g

~

060 M

r

{

rn

~

0.50 O

CI7 VP

(

w

~

040 I

0.30 o

I Pl 0.20 Vl C7 W 27 po Wt:

m o~

~ Vl Ulmo ZJ Q)0 ol

-~z~

o C7 C-l~

~m l/l I

O Vl-

%z

+o m~

C) mmo e

l m~

K ~ cl Vl Pl

,XI o

O.l0 0.00 l50 l20 90 60 30 0

l50 l20 90 60 30 0

CEA WlTHDRAWAL-INCHES

/REACTIVITY CONTROL SYSTEMS':

PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.7 The part length CEA groups shall be limited to the insertion limits shown on Figure 3.1-5 with PLCEA insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to:

a.

< 7 EFPD per 30 EFPD interval, and b.

< 14 EFPD per calender year.

APPLICABILITY:

MODELS 1" and 2*

ACTION:

a.

With the part length CEA groups inserted beyond the Transient Insertion Limit, except for surveillance testing pursuant to Specification 4. l. 3. 1. 2, within two hours, either:

1.

Restore the part length CEA group to within the limits, or 2.

Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group position using Figure 3.1-5.

b.

With the part length CEA groups inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit for intervals

> 7 EFPD per 30 EFPD interval or > 14 EFPD per calendar year, either:

1.

Restore the part length group within the Long Term Steady State Insertion Limit within two hours, or 2.

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 UIREMENTS 4.1.3.7 The position of the part length CEA group shall be determined to be within the Transient Insertion Limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE " UNIT 1 3/4 1-33 AMENDMENT NO.

24

'e Io C7m I

I.OO 0.90 CV I

I/I m

Al C)

IM Vl n

Ill 4)

IQI 0.80 n

0.70 o'Z 0.60 om 0.50 O.CO 0.30 o

0.20 IoX Vl lv o Vl W Vl nX-I VImm Vl gr o

oP on mn 27 m o+

om mVI CI vn om

+O Vl m

o CII O.IO 0.00 Ulo o

oo CO

.V CII gl a

CA M

o o

PART LENGTH CEA POSITION, INCHES WITHDRAWN

3/4. 2 POWER DISTRIBUTION LIMITS 3/4 2.1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate limit of 13.5 kW/ft shall be maintained by one of the following methods as applicable:,

a.

Maintaining COLSS calculated core power less than or equal to the COLSS calculated power operating limit based on linear heat rate (when COLSS is in service);

or b.

Maintaining peak linear heat rate with'.n its limit using any operable CPC channel (when COLSS is out of service).

APPLICABILITY:

MODE 1 above 20K of RATED THERMAL POWER.

ACTION:

With the linear heat rate limit not being maintained as indicated by:

COLSS calculated core power exceeding the COLSS calculated core power operating limit based on linear heat rate; or 2.

Peak linear heat rate outside its limit using any operable CPC channel (when COLSS is out of service);

within 15 minutes initiate corrective action to reduce the linear neat rate to within the limits and either:

Restore the linear heat rate to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or b.

Reduce THERMAL POWER to less than or equal to 2(C of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.1. 1 The provisions of Specification 4.0.4 are not applicable.

4. 2. 1. 2 The linear heat rate shall be determined to be within its limit when THERMAL POWER is above 20K of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that the linear heat rate, as indicated on any OPERABLE Local Power Density channel, is within its limit.

4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on linear heat rate.

PALO VERDE - UNIT 1 AMENDMENT NO. 24

POWER DISTRIBUTION LIMITS 3/4.2.2 PLANAR RADIAL PEAKING FACTORS -

FX LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F

) shall be less than or xy equal to the PLANAR RADIAL PEAKING FACTORS (F

) used in the Core Operating xy Limit Supervisory System (COLSS) and in the Core Protection Calculators (CPC).

APPLICABILITY:

MODE 1 above 20K of RATED THERMAL POWER.*

ACTION:

With an Fexceeding a corresponding F, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

a.

Adjust the CPC addressable constants to increase the multiplier applied to planar radial peaking by a factor equivalent to greater than or equal to F

/F and restrict subsequent operation so that a

xy xy margin to the COLSS operating limits of at least [(F /F ) - 1.0]

x 100K is maintained; or b.

Adjust the affected PLANAR RADIAL PEAKING FACTORS (F ) used in the xy COLSS and CPC to a value greater than. or equal to the measured PLANAR RADIAL PEAKING FACTORS (F ) or c.

Reduce THERMAL POWER to less than or equal to 20X of RATED THERMAL POWER.

SURVEILLANCE RE UIREMENTS

4. 2. 2. 1 The provisions of Specification 4. 0.4 are not applicable.

4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F ) obtained by using xy the incore detection

system, shall be determined to be less than or equal to the PLANAR RADIAL PEAKING FACTORS (F

), used in the COLSS and CPC at the xy

'ollowing intervals:

a.

After each fuel loading with THERMAL POWER greater than 40K but prior to operation above 70K of RATED THERMAL POWER, and b.

At least once per 31 Effective Full Power Days.

See Special"Test Exception 3. 10.2.

PALO VERDE - UNIT 1 3/4 2-2

POWER DISTRIBUTION LIMITS 3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by one of the following methods:

a.

Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR (when COLSS is in service, and either one or both CEACs are operable);

or b.

Maintaining COLSS calculated core power less than or equal to COLSS calculated core power operating limit based on DNBR decreased by the allowance shown in Figure 3.2-1 (when COLSS is in service and neither CEAC is operable);

or c.

Operating within the region of acceptable operation of Figure 3.2-2 using any operable CPC channel (when COLSS is out of service and either one or both CEACs are operable);

or d.

Operating within the region of acceptable operation of Figure 3.2-2A using any operable CPC channel (when COLSS is out of service and neither CEAC is operable).

APPLICABILITY:

MODE 1 above 20K of RATED THERMAL POWER.

ACTION:

With the DNBR not bein maintained:

2.

As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or With COLSS out of service, operation outside the region of acceptable operation of Figure 3.2-2 or 3.2-2A, as applicable; within 15 minutes initiate corrective action to increase the DNBR to within the limits and either:

a.

Restore the DNBR to within its limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , or b.

Reduce THERMAL POWER to less than or equal to 20K of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.

4 ~ 2.4.2 The DNBR shall be determined to be within its limits when. THERMAL POWER is above 20X of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System

(COLSS) or, with the COLSS out of service, by verifying at least once per hours that the DNBR, as indicated on any OPERABLE DNBR channel, is within the limit shown on Figure 3.2-2 or Figure 3.2-2A.

4.2.4.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a.THERMAL POWER level less than or equal to the core power

, operating limit based on DNBR.

PALO VERDE - UNIT 1 3/4 2-5 AMENDMENT NO.

24

25 I

Cl I

O Or

~ CA W Ill M C) mM Tl K7 Kl OO 07 ~

Om mm OQ Ul Z A O~t tel M K7 m CD W Im M

R7m n

C)

CA CA CJ oN O

O

~o C3 Pl Pl 3>

Q I

C3 Pl C)n C) t'5 28% c 95%

POWER 75 188 PERCENT OF RATEO THERMAL POWER

COLSS OUT OF SERVICE DNBR LIMIT LINE 2.1 2.8 ACCEPTABLE OPERATION MINIMUM 1 CEAC OPERABLE 1.8

(.1,1.85)

(.2,1.85) 1.7

(-.2,1.75)

UNACCEPTABLE OPERATION 1.6 1.5

-8.3

-8.2

-8.1 8.1 8.2 8.3 CORE AVERAGE ASI FIGURE 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE)

PALO VERDE - UNIT 1 3/4 2-7 AMENDMENT NO.

24

COLSS OUT OF SERVICE DNBR LIMIT LINE 2.3 ACCEPTABLE OPERATION CEACs INOPERABLE

( 85 2 38)

(.2,2.38)

CL CO 0

7 2 2.1 CL 2.8

(-.2,2.13)

UNACCEPTABLE OPERATION 1.9-8.3

-8.2

-8.1 8.1 8.2 8.3 CORE AVERAGE ASI FIGURE 3.2-2A DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE)

PALO VERDE UNIT 1 3/4 2-7a AMENDMENT NO.

24

POWER DISTRIBUTION LIMITS 3/4.2.5 RCS FLOW RATE LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reactor Coolant System total flow rate shall be greater than or equal to 155.8 x 106 ibm/hr.

APPLICABILITY:

MODE 1.

ACTION:

With the actual Reactor Coolant System total flow. rate determined to be less than the above limit, reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than or equal to its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

PALO VERDE - UNIT 1 3/4 2-8 AMENDMENT NO 24

gI 1

t t I

POWER DISTRIBUTION LIMITS 3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION C

3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a.

COLSS OPERABLE

-0. 28

< ASI < 0. 28 b,

COLSS OUT OF SERVICE (CPC)

-0. 20

< ASI < + 0.20 APPLICABILITY:

MODE 1 above 20K of RATED THERMAL POWER".

ACTION:

With the core average AXIAL SHAPE INDEX outside its above limits, restore the core average ASI to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 20X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using the COLSS or any OPERABLE Core Protection Calculator channel.

See Special Test Exception 3. 10.2.

PALO VERDE - UNIT 1 3/4 2-11

POWER DISTRIBUTION LIMITS 3/4.2.8 PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The pressurizer pressure shall be maintained between 2025 psia and 2300 psia.

APPLICABILITY:

MODES 1 and 2."

ACTION:

With the pressurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS

4. 2. 8 The pressurizer pressure shall be determined to be within its limit at

'least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

"See Special Test Exception 3. 10.5 PALO VERDE - UNIT 1 3/4 2-12 AMENDMENT NO. 24

TABLE 3 ~ 3-1. (Continued)

ACTION STATEMENTS ACTION 4 ACTION 5 ACTION 6 4.

Steam Generator Level -

Low (Wide Range) 3.

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

Steam Generator Level 2-Low (ESF)

Steam Generator Level - Low (RPS)

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 requirement, 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, 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.

a.

With one CEAC inoperable, operation may continue for up to 7 days provided that at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , each CEA is verified to be within 6.6 inches (indicated position) of all other CEAs in its group.

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 1 3/4 3-7 AMENDMENT NO. 24

TABLE 3. 3-1 Continued ACTION STATEMENTS 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 are withdrawn to and subsequently maintained at the "Full Out" position, except during surveillance testing pursuant to the requirements of Specifica-tion 4."1.3.1.2 or for control when CEA group 5

may be inserted no further than 127.5 inches withdrawn.

ACTION 7 ACTION 8 b)

The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to be indicated that both CEAC's are inoperab'le.

c)

The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the "Standby" mode except during CEA group 5 motion permitted by a) above, when the CEDMCS may be operated in either the "Manual Group" or "Manual Individual" mode.

3.

At least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , all fu11-length and part-length CEAs are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3. 1.2 or during insertion of CEA group 5 as permitted by 2. a) above, then verify at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned within 6.6 inches (indicated position) of all other CEAs in its group.

With three or more auto restarts, excluding periodic auto restarts (Code 30 and Code 33), of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a

CHANNEL FUNCTIONAL TEST within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel to OPERABLE 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 hour.

PALO VERDE - UNIT 1 3/4 3"8 AMENDMENT NO ~ 24

THIS PAGE INTENTIALLY LEFT BLANK PALO VERDE " UNIT 1 3/4 3-9 AMENDMENT NO. 24

wf

~ ~

THIS PAGE INTENTIALLY LEFT BLANK PALO VERDE - UNIT 1 3/4 3-10 AMENDMENT NO. ~4

TABLE 3.3-2 CD(

FUNCTIONAL UNIT I.

TRIP GENERATION A.

Process 1.

Pressurizer Pressure

- High 2.

Pressurizer Pressure

- Low

RESPONSE

TIME

< 1.15 seconds

< 1. 15 seconds

< 1.15 seconds

< 0.58 second 3.

Steam Generator Level - Low Steam Generator Level - High Steam Generator Pressure - Low 5.

Containment Pressure High Reactor Coolant Flow - Low 6.

7.

REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES 8.

9.

Local Power Density - High a.

Neutron Flux Power from Excore Neutron Detectors b.

CEA Positions c.

CEA Positions:

CEAC Penalty Factor DNBR - Low

< 0.75 second~

< 1.35 second""

< 0.75 second"*

a@

b.

C.

d.

e.f.

g.

Neutron Flux Power from Excore Neutron Detectors CEA Positions Cold Leg Temperature Hot Leg Temperature Primary Coolant Pump Shaft Speed Reactor Coolant Pressure from Pressurizer CEA Positions:

CEAC Penalty Factor

< 0.75

< 1.35

< 0.75

< 0.30

< 0.75

< 0.75 second" second**

second¹¹ second¹¹ second¹ second¹¹¹ second*"

CD m

B.

Excore Neutron Flux

< 0.55 second" 2.

Variable Overpower Trip Logarithmic Power Level - High a.

Startup and Operating b.

Shutdown

< 0.55 second~

< 0.55 second"

TABLE 3.3-2 (Continued)

C)(

m FUNCTIONAL UNIT I

C.

V

RESPONSE

TIME Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators Supplementary Protection System Pressurizer Pressure - High Not Applicable Not Applicable D.

< 1.15 second II.

RPS LOGIC REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES A.

Matrix Logic B.

Initiation Logic III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers B.

Manual Trip Not Applicable Not Applicable Not Applicable Not Applicable Neutron detectors are exempt from response time testing.

The response time of the neutron flux signal portion of the channel shall be measured from the detector output or from the input of first electronic component in channel.

Response

time shall be measured from the output of the sensor.

Acceptable CEA sensor response shall be demonstrated by compliance with Specification 3.1.3.4.

¹The pulse transmitters measuring.

pump speed are exempt from response time testing.

The response time shall be measured from the pulse shaper input.

¹¹Response time shall be measured from the output of the resistance temperature detector (sensor).

RTD response time shall be measured at least once per 18 months.

The measured response time of the slowest RTD shall be less than or equal to 8 seconds.

¹¹¹Response time shall be measured from the output of the pressure transmitter.

The transmitter response time shall be less than or equal to 0.7 second.

THIS PAGE INTENTIALLYLEFT BLANK PALO VERDE - UNIT 1 3/4 3-X3 AMENDMENT NO. 24

TABLE 4. 3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT I.

TRIP GENERATION A.

Process CHANNEL CHANNEL CHECK CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE RE UIRED 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.

DNBR - Low B.

Excore Neutron Flux S

S S

S

~ S R

R R

R

,R 0 (2, 4),

R (4, 5)

D (2, 4),

R (4, 5)

M (8),

S (7)

M M

M M,

R (6)

M, R (6) 1, 2

1, 2

1 2

1.

2 3A 4*

1, 2

1 2

1.

Variable Overpower Trip 2.

Logarithmic Power Level - High C.

Core Protection Calculator System 1.

CEA Calculators 2.

Core Protection Calculators D (2, 4),

M (3, 4) 0 (4)

R (4)

R D {2, 4),

R {4, 5)

M (8),

S (7) 1, 2

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

and "

M, R(6) 1,2 M {9),

R {6) 1, 2

TABLE 3.3-4 ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ESFA SYSTEM FUNCTIONAL UNIT I.

SAFETY INJECTION (SIAS)

A.

Sensor/Trip Units 1.

Containment Pressure

- High 2.

Pressurizer Pressure

- Low B.

ESFA System Logic C.

Actuation Systems II.

CONTAINMENT ISOLATION (CIAS)

A.

Sensor/Trip Units I.

Containment Pressure

- High 2.

Pressurizer Pressure - Low B.

ESFA System Logic C.

Actuation Systems III. CONTAINMENT SPRAY (CSAS)

A.

Sensor/Trip Units Containment Pressure High - High B.

ESFA System Logic C.

Actuation Systems IV.

MAIN STEAM LINE ISOLATION (MSIS)

A.

Sensor/Trip Units 1.

Steam Generator Pressure Low 2.

Steam Generator Level - High 3.

Containment Pressure

- High B.

ESFA System Logic C.

Actuation Systems TRIP SETPOINT

< 3.0 psig

> 1837 psia Not Applicable Not Applicable

< 3.0 psig

> 1837 psia Not Applicable Not Applicable

< 8.5 psig Not Applicable Not Applicable

> 9>9 psia(3)

< 91.0X NR(')

< 3.0 psig Not Applicable Not Applicable ALLOWABLE VALUES 3.2 pslg

> 1822 ps'a(

)

Not Applicable Not Applicable

< 3.2 psig.

> 1822 psia(

)

Not Applicable Not Applicable

< 8 9 psig Not Applicable Not Applicable

> 912 psia

< 91..5X NR(2)

3. 2 ps1g Not Applicable Not Applicable

I/

TABLE 3.3-4 (Continued)

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

RECIRCULATION (RAS)

A.

Sensor/Trip Units Refueling Water Storage Tank - Low B.

ESFA System Logic C.

Actuation System VI.

AUXILIARY FEEDWATER (SG-1) (AFAS-1)

A.

Sensor/Trip Units 1.

Steam Generator ¹1 Level - Low 2.

Steam Generator b, Pressure-SG2

> SG1 B.

ESFA System Logic C.

Actuation Systems VII. AUXILIARYFEEDWATER (SG-2)(AFAS-2)

A.

Sensor/Trip Units 1.

Steam Generator

¹2 Level - Low 2.

Steam Generator b, Pressure-SG1 >

SG2 B.

ESFA System Logic C.

Actuation Systems VIII.

LOSS OF POWER A.

4.16 kV Emergency Bus Undervoltage (Loss of Voltage)

B.

4. 16 kV Emergency Bus Undervoltage (Degraded Voltage)

IX.

CONTROL ROOM ESSENTIAL FILTRATION TRIP VALUES 7.4X of Span Not Applicable Not Applicable

> 25.8X WR

< 185 psid Not Applicable Not Applicable

> 25.8X WR( )

< 185 psid Not Applicable Not Applicable

> 3250 volts 2930 to 3744 volts with a 35-second maximum time delay

< 2 x 10-s pCi/cc ALLOWABLE VALUES 7.9

> X of Span

> 6.9 Not Applicable Not Applicable

> 25.3X WR( )

< 192 psid Not Appli cable Not Applicable

> 25.3X WR

< 192 psid Not Applicable Not Applicable

> 3250 volts 2930 to 3744 volts with a 35-second maximum time delay

< 2 x 10-s pCi/cc

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES Continued 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 ld greater than or cold equal to 552'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 fry 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.

7o 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 aff'ected 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 per turbations 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 1 B 3/4 1-5

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES Continued) and load maneuvering.

Analyses are performed based on the 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 used in the analyses and provide a means of preserving the assumptions on CEA insertions used.

The limits speci-fied serve to limit the behavior of the radial peaking factors within the bounds determined from analysis.

The actions specified serve to lim'it the extent of radial xenon redistribution effects to those accommodated in the analyses.

The Long and Short Term Insertion Limits of Specifications

3. 1.3.6 and 3.1.3.7 are specified for 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 Insertion Limits of Specifications 3.1.3.6 and 3. 1.3.7 and the Shutdown CEA Insertion Limits of Specification

3. 1. 3. 5 ensure that (1) the minimum SHUTDOWN MARGIN is maintained, and (2) the potential effects of a CEA ejection accident are limited to acceptable levels.

Long-term operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial peaking factors.

The PVNGS CPC and COLSS systems are responsible for the safety and monitoring functions, respectively, of the reactor core.

COLSS monitors the DNB Power Operating Limit (POL) and various operating par'ameters to help the operator main-tain plant operation within the limiting conditions for operation (LCO).

Operat-ing within the LCO guarantees that in the event of an Anticipated Operational Occurrence (AOO), the CPCs will provide a reactor trip in time to prevent un-acceptable fuel damage.

The COLSS reserves the Required Overpower Margin (ROPM) to account for the Loss of Flow (LOF) and CEA misoperation transients.

When the COLSS is Out of Service (COOS), the monitoring function is performed via the CPC calculation of DNBR in conjunction with Technical Specification COOS Limit Lines (Figures 3'-2 and 3.2-2A) which restrict the reactor power sufficiently to preserve the ROPM.

The reduction of the CEA deviation penalties in accordance with the CEAC (Control Element Assembly Calculator) sensitivity reduction program has been performed.

This task involved 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 of the CEA deviation penalty in either the CPC DNB and LHR (Linear 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 8 3/4 1-6 AMENDMENT NO. 24

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued)

If an inward CEA deviation event occurs, the current CPC algorithm applies two penalty factors to each of the DNB and LHR calculations.

The first, a static penalty factor, is applied upon detection of the event.

The second, a xenon redistribution penalty, is applied linearly as a function of time after the CEA drop.

The expected margin degradation for the inward CEA deviation event for which the penalty factor has been reduced is accounted for in two ways.

The ROPM reserved in COLSS is used to account for some of the margin degrada-tion.

Further, a power reduction in accordance with the curve in Figure 3. 1-2A is required.

In addition, the part length CEA maneuvering is restricted in accordance with Figure 3. 1-5 to justify reduction of the PLR deviation penalty factors.

The technical specification permits plant operation if both CEACs are considered inoperable for safety purposes after this period.

PALO VERDE UNIT 1 B 3/4 1-7 AMENDMENT NO.

I

3/4. 2 POWER DISTRIBUTION LIMITS BASES 3/4. 2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a

LOCA, the peak temperature of the fuel cladding will not exceed 2200~F.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate.

Reactor operation at or below this calculated power level assures that the limits of 13.5 kW/ft are not exceeded.

The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator.

A COLSS alarm is annunciated in the event that the core power exceeds the core power operating limit.

This provides adequate margin to the linear heat rate operating limit for normal steady-state opera-tion.

Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded.

In the event this occurs, COLSS, alarms will be annunciated.

If the event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation.

The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability/

confidence level, that the maximum linear heat rate calculated by COLSS is conservative with respect to the actual maximum linear heat rate existing in the core.

These penalty factor s are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux uncertainty, axial densification, software algorithm modelling, computer processing, rod bow, and core power measurement.

Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB, and total core power are also monitored by the CPCs.

Therefore, in the event that the COLSS is not being used, operation within the linear heat rate limit can be maintained by utilizing any operable CPC channel.

The above listed uncertainty and penalty factors plus those asso-ciated with the CPC startup test acceptance criteria are also included in the CPCs.

PALO VERDE - UNIT 1 B 3/4 2-1 AMENDMENT NO. 24

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 PLANAR RADIAL PEAKING FACTORS Limiting the values of the PLANAR RADIAL PEAKING FACTORS (Fx ) used in the COLSS and CPCs to values equal to or greater than the measured PLANAR RADIAL PEAKING FACTORS (F

) provides assurance that the limits calculated by COLSS xy and the CPCs remain valid.

Data from the incore detectors are used for determining the measured PLANAR RADIAL PEAKING FACTORS.

A minimum core. power at 20K of RATED THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS.

The 20X RATED THERMAL POWER threshold is due to the neutron flux detector system being inaccurate below 20K core power.

Core noise level at low power is too large to obtain usable detector readings.

The periodic surveillance requirements for determining the measured PLANAR RADIAL PEAKING FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle.

Determining the measured PLANAR RADIAL PEAKING FACTORS after each fuel loading prior to exceeding 70K of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

3/4.2.3 AZIMUTHAL POWER TILT - T The limitations on -the AZIMUTHAL POWER TILT are provided to ensure that design safety margins are maintained.

An AZIMUTHAL POWER TILT greater than

0. 10 is not expected and if it should occur, operation is restricted to only those conditions required to identify the cause of the tilt.

The tilt is normally calculated by COLSS.

A minimum core power of 20X of RATED THERMAL POWER is assumed by the CPCs in its input to COLSS for calculation of AZIMUTHAL POWER TILT.

The 20X RATED THERMAL. POWER threshold is due to the neutron flux detector system being inaccurate below 20K core power.

Core noise level at low power is too large to obtain usable detector readings.

The surveillance requirements specified when COLSS is out of service provide an acceptable means of detecting the presence of a steady-state tilt. It is necessary to explicitly account for power asymmetries because the radial peaking factors used in the core power distribution calculations are based on an untilted power distribution.

The AZIMUTHAL POWER TILT is equal to Ptilt untilt 1 0 where:

AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the untilted power at the location is of the form:

t lt t'lt g

tilt untilt q

where:

T is the peak fractional tilt amplitude at the core periphery q

g is the radial normalizing factor 6 is the azimuthal core location 6o ".is the azimuthal core location of maximum tilt PALO VERDE - UNIT 1 B 3/4 2-2

POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - T (Continued) q unt,-lt is the ratio of the power at -a core location in the presence of a tilt to the power at that location with no tilt.

The AZIMUTHAL POWER TILT allowance used in the CPCs is defined as the value of CPC addressable constant TR-1.0.

3/4.2.4 DNBR MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a

conservative envelope of operating conditions consistent with the safety analy-sis assumptions and which have been analytically demonstrated adequate to main-tain an acceptable minimum DNBR throughout all anticipated operational occur-rences.

Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the DNBR channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR.

The COLSS calculation of core power operating limit based on DNBR includes appropriate penalty factors which provide, with a 95/95 probability/confidence level, that the core power limits calculated by COLSS (based on the minimum DNBR Limit) are conservative with respect to the actual core power limit.

These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux, state parameter measurement, software algorithm modelling, computer processing, rod

bow, and core power measurement.

Parameters required to maintain the margin to DNB and total core power are also monitored by the CPCs.

Therefore, in the event that the COLSS is not being used, operation within the limits of Figures 3.2-2 and 3.2-2A can be maintained by utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels.

The above listed uncertainty and penalty factors are also included in the CPCs which assume a minimum core power of 20K of RATED THERMAL POWER.

The 20K RATED THERMAL POWER threshold is due to the neutron flux detector system being less accurate below 20K core power.

Core noise level at low power is too large to obtain usable detector readings.

A DNBR penalty factor has been included in the COLSS and CPC DNBR calcu-lations to accommodate the effects of rod bow.

The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly.

Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow.

Conversely, lower burnup assemblies will experience less rod bow.

In design calculations, the penalty for each batch required to com-pensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak.

A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.

PALO VERDE - UNIT 1 B 3/4 2-3 AMENDMENT NO. 24

POWER DISTRIBUTION LIMITS BASES 3/4.2 '

RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate is maintained at or above the minimum value used in the safety analyses.

The minimum value used in the safety analyses is 95K of the design flow rate (164.0 x 10 ibm/hr) or 155.8 x 10 ibm/hr.

The actual RCS flow rate is deter-mined by direct measurement and an uncertainty associated with that measurement is consider ed when comparing actual RCS flow rate to the minimum required value of 155. 8 x 10'bm/hr.

3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses.

3/4. 2.7 AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of the core average AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.

3/4.2.8 PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.

PALO VERDE - UNIT 1 B 3/4 2"4 AMENDMENT NO. 24

3/4.3 INSTRUMENTATION BASES 3/4.3.

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

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

Response

time testing of resistance temperature

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

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

Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3. 1 and 6.8. 1) ensure that inadvertent misloading of addressable constants into the CPCs is unlikely.

Any modifications which are made to the core protection calculator soft-ware (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with "CPC Protection Algorithm Software Change Proce-dure," CEN-39(A)-P, Revision 3-P-A and Supplement 1-P, Revision 3-P-A or another NRC approved procedure on CPC software modifications.

CPC modifications which result in a) an unreviewed safety questions, b) a Technical Specification change, or c) methodology not previously approved by the NRC, including additions or deletions to addressable constants or modifi-cations to the approved constant limit values, will require NRC approval prior to implementation.

The design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable.

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

If the second CEAC fails, the CPCs in conjunction with plant Technical Specifications will use DNBR and LPD penalty factors and increased DNBR and LPD margin to restrict reactor operation to a power level that will ensure safe operation of the plant. If the margins are not maintained, a reactor trip will occur.

PALO VERDE - UNIT l B 3/4 3-1 AMENDMENT NO. 24

INSTRUMENTATION BASES REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Continued The value of the DNBR in Specification

2. 1 is conservatively compensated for measurement uncertainties.

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

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

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

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

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

24

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3. 1 The reactor core shall contain 241 fuel assemblies with each fuel assembly containing 236 fuel rods or'urnable poison rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of l50 inches and contain a maximum total weight of approximately'950 grams uranium.

Each burnable poison rod shall have a nominal active poison length of 136 inches..

The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.05 weight percent U-235.

'ONTROL ELEMENT ASSEMBLIES 5.3.2 The reactor core shall contain 76 full-length and 13 part-length control element assemblies.

5 4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 a.

b.

C.

The Reactor Coolant System is designed and sha'll be maintained:

In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable surveillance requirements, For a pressure of 2500 psia, and For a temperature of 6500F, except for the pressurizer which is 700 F

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 13,900

+ 300/-0 cubic feet at a nominal T

of 593'F.

PALO VERDE - UNIT 1 5"5 AMENDMENT NO. 24

DESIGN FEATURES

5. 5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5. 6 FUEL STORAGE

5. 6. 1 CRITICALITY 5.6.1. 1 The spent fuel storage racks are designed and shall be maintained with:

a.

A k equivalent to less than or equal to 0.95 when flooded with unblotted water, which includes a conservative allowance of 2.6X delta k/k for uncertainties as described in Section

9. 1 of the FSAR.

b.

A nominal 9.5 inch center-to-center distance between fuel assemblies placed in the storage racks in a high density configuration.

5.6.1.2 The k ff for new fuel for the first core loading stored dry in the eff spent fuel storage racks shall not exceed 0.98 when aqueous foam moderation is assumed.

DRAINAGE 5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 137 feet - 6 inches.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1329 fuel assemblies.

5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Tables 5.7-1 and 5.7-2.

PALO VERDE - UNIT 1

ML17300B099

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