Text

Amend 19 to License NPF-51,revising Several Portions of Tech Specs,App a in Support of Cycle 2 Operation
	ML20154E773
Person / Time
Site:	Palo Verde
Issue date:	05/05/1988
From:	Knighton G; Office of Nuclear Reactor Regulation
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UNITED STATES 1

NUCLEAR REGULATdRY COMMISSION g

i WASHINGTON, D. C. 20666

\\...../

ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-529 PALO VERDE NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.19 License No. NPF-51 1.

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

A.

The application for amendment, dated December 3,1987, as supplemented by letters dated December 2, 1987 and February 4, 1988, 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 Conipany, Public j

Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority (licensees),

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

The facility will operate in confonnity with the application, the provisions of Act, and the regulations of the Comission; 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon 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 Comission'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 enclosure to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-51 is hereby amended to read as folicws:

8805200323 080505 DR ADOCK 0500 9

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

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.

3.

This license amendment is effective as of the date of issuance. The changes in the Technical Specifications are to become effective within 30 days of issuance of the amendment.

In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized.

FOR THE NUCLEAR REGUL TORY COMMISSION George W' Knighton trector Project Directorate V Division of Reactor Projects - III, IV, V and Spe.cial Projects

Enclosure:

Changes to the Technical Specifications Date of Issuance: May 5, 1988

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May 5, 1988 ENCLOSURE TO LICENSE AMENDMENT AMENDMENT NO.19 TO FACILITY OPERATING LICENSE N0. NPF '

DOCKET NO. STN 50-529 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.

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

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

Arendment Pages Overleaf Pages

!V III XIX XX 2-1 2-2 2-3 2-4 2-5 2-6 B 2-1 B 2-2 B 2-3 B 2-4 8 2-5 B 2-6 3/4 1-2a 3/4 1-5 3/4 1-6 3/4 1-21 3/4 1-22 3/4 1-23 3/4 1-24 3/4 1-25 3/4 1-26 3/4 1-31 3/4 1-32 3/4 1-33 3/4 1-34 3/4 2-1 3/4 2-2 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-7a 3/4 2-B 3/4 2-12 3/4 2-11

DOCKET NO. STN 50-529 Amendment Pages Overleaf Pages 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 3-14 3/4 3-26 3/4 3-25 3/4 10-2 3/4 10-1 3/4 10-4 3/4 10-3 B 3/4 1-6 B 3/4 1-5 B 3/4 1-7 B 3/4 2-1 B 3/4 2-2 B 3/4 2-3 B 3/4 2-4 B 3/4 3-1 B 3/4 3-2 i

B 3/4 3-3 B 3/4 3-4 B 3/4 3-5 b

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I INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE.............................................

2-1 2.1.1.1 0NBR.....................................................

2-1 2.1.1.2 PEAK LINEAR HEAT RATE....................................

2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE..........................

2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETP01NTS.....................................

2-2 j

BASES SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE...............................................

B 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE............................

B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETP0INTS.....................................

B 2-2 9

PALO VERDE - UNIT 2 III

l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTI0h PAGE 3/4.0 APPLICABILITY..............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - ALL CEAs FULLY INSERTED.............

3/4 1-1 SHUTDOWN MARGIN - X

- ANY CEA WITH0RAWN............

3/4 1-2 N-1 MODERATOR TEMPERATURE C0 EFFICIENT.....................

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

3/4 1-6 3/4.1.2 BORATION SYSTEMS FLOW PATHS - SHUT 00WN.................................

3/4 1-7 FLOW PATHS - 0PERATING...............................,

3/4 1-8 CHARGING' PUMPS - SHUT 00WN.............................

3/4 1-9 CHARGING PUMPS - 0PERATING............................

3/4 1-10 B0 RATED WATER SOURCES - SHUT 00WN......................

3/4 1-11 BORATED WATER SOURCES - OPERATING.....................

3/4 1-13 BORON DILUTION ALARMS.................................

3/4 1-14 3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA P0SITION..........................................

3/4 1-21 POSITION INDICATOR CHANNELS - OPERATING...............

3/4 1-25 POSITION INDICATOR CHANNELS - SHUT 00WN................

3/4 1-26 CEA OROP TIME.........................................

3/4 1-27 SHUTDOWN CEA INSERTION LIMIT..........................

3/4 1-28 REGULATING CEA INSERTION LIMITS.......................

3/4 1-29 PART LENGTH CEA INSERTION LIMITS......................

3/4 1-33 PALO VERDE - UNIT 2 IV AMEN 0 MENT NO. 19

j

~

a.

INDEX LIST OF FIGURES PAGE 3.1-1A SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............

3/4 1-2a 3.1-1 ALLOWABLE MTC MODES 1 AND 2............................

3/4 1-5 3.1-2 MINIMUM BORATED WATER V0LUMES..........................

3/4 1-12 3.1-2A CORE POWER LIMIT AFTER CEA DEVIATION...................

3/4 1-24 3.1-3 CEA INSERTION LIMITS VS THERMAL POWER (COLSS IN SERVICE)............................'.........

3/4 1-31 3.1-4 CEA INSERTION LIMITS VS THERMAL POWER (COLSS OUT OF SERVICE).................................

3/4 1-32 3.1-5 PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER.......

3/4 1-34 3.2-1 COLSS ONBR POWER OPERATING LIMIT ALLOWANCE FOR BOTH CEACs IN0PERABLE.......................................

3/4 2-6 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE).....

3/4 2-7 3.2-2A DNBR MARGIN OPERATING LIMIT BASED ON CORE. PROT.ECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE)...

3/4 2-7a 3.2-3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVEL..................................................

3/4 2-10 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY

> 1.0 pCi/ GRAM DOSE EQUIVALENT I-131...................

3/4 4-27 3.4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 0 TO 10 YEARS OF FULL POWER OPERATION..............................................

3/4 4-29 4.7-1 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..............

3/4 7-26 8 3/4.4-1 NIL-00CTILITY TRANSITION TEMPERATURE INCREAfE AS A FUNCTION OF FAST (E > 1 MeV) NEUTRON FLUENCE (550 F IRRADIATION)....................................

B 3/4 4-10 5.1-1 SITE AND EXCLUSION BOUNDARIES..........................

5-2 5.1-2 LOW POPULATION Z0NE....................................

5-3 5.1-3 GASEOUS RELEASE P0lNTS.................................

5-4 6.2-1 0FFSITE ORGANIZATION..................................

6-3 6.2-2 ONSITE ORGANIZATION...................................

6-4 PALO VERDE - UNIT 2 XIX AMENOMENT NO. 19

INDEX LIST OF TABLES PAGE 1.1 FREQUENCY N0TATION......................................

1-8 1.2 OPERATIONAL M0 DES.......................................

1-9 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS..................................................

2-3 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL M0 DES.....................

3.1-1 FOR K,ff > 0.98.........................................

3/4 1-16 3.1-2 FOR 0.98 > K,ff > 0.97..................................

3/4 1-17 3.1-3 F O R 0. 9 7 > K, f f > 0. 9 5..................................

3/4 1-18 3.1-4 F O R 0. 9 6 > K, f f > 0. 9 5.................................

3/4 1-19 3.1-5 FOR K 1 0.95.........................................

3/4 1-20 eff 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION......................

3/4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.......

3/4 3-11 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-14

'3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................

3/4 3-18 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES.............................

3/4 3-25 3.3-5 ENGI'HEERED SAFETY FEATURES RESPONSE TIMES...............

3/4 3-28 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-31 3.3-6 RADIATION MONITORING INSTRUMENTATION....................

3/4 3-38 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-40 3.3-7 SEISMIC MONITORING INSTRUMENTATION......................

3/4 3-43 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................

3/4 3-44 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION...............

3/4 3-46 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-47 3.3-9A REMOTE SHUTDOWN INSTRUMENTATION.........................

3/4 3-49 3.3-98 REMOTE SHUTDOWN DISCONNECT SWITCHES.....................

3/4 3-50 PALO VERDE - UNIT 2 XX AMENDMENT NO. 19

2.0 SAFETY LIMIT < AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ONBR 2.1.1.1 The calculated DNBR of the reactor core shall be maintained greater j

than or equal to 1.24.

i APPLICABILITY:

MODES 1 and 2.

ACTIO'h 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:

Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT STAN0BY 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 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, MODES 3, 4, and 5:

Whenever 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 2 2-1 AMENDMENT N0. 19

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.

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 Satpoint value.

I I

PALO VERDE - UNIT 2 2-2

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TABLE 2.2-1 mN o

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LINITS m8 m

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

[

I.

TRIP GENERATION h

A.

Process m

1.

Pressurizer Pressure - High 5 2383 psia

$ 2388 psia 2.

Pressurizer Pressure - Low 1 1037 psia (2) 3 1822 psia (2) 3.

Steam Generator Level - Low 2 44.2% (4) 1 43.7% (4) 4.

Steam Generator Level - High 5 91.0% (9)

.1 91.5% (9) 5.

Steam Generator Pressure - Low 3 919 psia (3)

> 912 psia (3) 6.

Containment Pressure - High 5 3.0 psig 5 3.2 psig 7.

Reactor Coolant Flow - Low 0

a.

Rate 1 0.115 psi /sec (6)(7)

$ 0.118 psi /sec (6)(7) b.

Floor 1 11.9 psid(6)(7) 1 11.7 psid (6)(7) c.

Band

$ 10.0 psid(6)(7)

$ 10.2 psid (6)(7) 8.

Local Power Density - High 1 21.0 kW/ft (5)

$ 21.0 kW/ft (5) 9.

DNBR - Low 1 1.24 (5) 1 1.24 (5)

B.

Excore Neutron Flux 1.

Variable Overpower Trip a.

Rate

'< 10.6%/ min of RATED

< 11.0%/ min of RATED g

THERMAL POWER (8)

E b.

Ceiling

< 110.0% of RATED

< 111.0% of RATED THERMAL POWER (8)

THERMAL POWER (8) 5 c.

Band

< 9.8% of RATED

< 10.0% of RATED g

THERMAL POWER (8)

G d

m

TABLE 2.2-1 (Continued)~

5 REACTOR PROTECTIVE INSTRtiMENTATION TRIP SETPOINT LIMITS 5

FUNCTIONAL UNIT

. TRIP SETPOINT Alt 0WA8tE VALUES 2.

Logarithmic Power Level - High (1) l a.

Startup and Operating

< 0.010% of RATE 0

< 0.011% of RATED THERMAL POWER THERMAL POWER b.

Shutdown"

< 0.010% of RATED

< 0.011% of RATED l

u THERMAL POWER THERMAL POWER

~

C.

Core Protection Calculator System 1.

CEA Calculators Not Appilcable Not Applicable 2.

Core Protection Calculators Not Appilcable Not Applicable N

D.

Supplementary Protection System a

Pressurizer Pressure - High

$ 2409 psia 5 2414 psia II.

RPS LOGIC A.

Matrix Logic Not Appilcable Not Applicable 8.

Initiation Logic Not.Appilcable Not Applicable 1

III. RPS ACTUATION DEVICES A.

Reactor Trip Breakers Not Appilcable Not Applicable 8.

Manulll Trip Not Applicable Not Applicable ~

,k

=

1 2

5 t

TABLE 2.2-1 (Continued) l REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 1

, TABLE NOTATIONS (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 4% 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 1

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 me.rgin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shell be increased automatically as steam generator pressure is increased until the trip setpoint is reached.

(4) % 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

'tha trip setpoint includes measurement, calculational and processor uncer,

tainties.

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

I 1

PALO VERDE - UNIT 2 2-5 AMENDMENT NO 19 I

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 restrictions on the rate at which the setpoint can in&rease.

FLOOR is the minimum value of the trip setpoint.

IIs3 is the amount by which the trip setpoint is below the input signal unless limited by Rate or Floor.

Setpoints are based on steam generator differential pressuke.

(7) The setpoint say 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 is the amount by which the trip setpoint is above the steady state input cignal unless limited by the rata or the ceiling.

i I

(9) % of the distance between steam generator upper and lower level narrow range instrument nozzles.,

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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 ccolant s&turation 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 maxim',o clad surface temperature is only slightly greater than the coolant saturation temperature.

The upper boundary of the nucleate boiling regi.me 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 ::ladding 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 I

of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.

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

Second, operation with a peak linear heat rate below that which would cause fuel centerline 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 2 B 2-1 AMENDMENT NO. 19

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 whicn permits a maximum transient pressure of 110% (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 Setpnints 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 betwen each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Th( DNBR - Low and Local Power Density - High are digitally gene ated j

trip setpoints based on Safety Limits of 1.24 and 21 kW/ft, respectively.

l 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 2 8 2-2 AMENOMENT NO. 19

BASES l

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

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

This trip function will trip the reactor when the indicated neutron flux power exceeds either a rate limited setpoint at a great enough rate or reaches a preset ceiling.

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

These trip satpoints are provided in Table 2.2-1.

Logarithmic Power level - High The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant Systes 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 j

trip when the THERMAL POWER level is above 10 4% of RATED THERMAL POWER; this bypass it automatically removed when the THERMAL POWER level decreases to 10 4% of RATED THERMAL POWER.

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

This trip's setpoint is below the nominal lift setting of the pressurizer safety valves and its operation minimizes the undesirable opera-tion of the pressurizer safety valves.

Pressurizer Pressure - Low The Presserizer 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 i

9 PALO VERDE - UNIT 2 B 2-3 AMENOMENT NO. 19

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

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

j 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 nomal 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 Oensity - High 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 powsr density is calculated in the reactor protective system utilizing the following information:

PALO VERDE - UNIT 2 B 2-4

c

/

= - -

BASES Local Power Density - High (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 local 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 LP0 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 censity), sensor time delays, and protection >ystem equipment time delays.

DNBR - Low The ONBR - Low trip is provided to prevent the ONBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences.

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

At this pressure l

a ONBR - 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.

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

Reactor coolant mass flow rate from reactor coolant pump speed; f.

Core inlet temperature from reactor coolant cold leg temperature measurements.

PALO VERDE - UNIT 2 B 2-5 AMEN 0 MENT NO. 19

SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES ONBR - Low (Continued)

The ONBR, 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 ONBR is sufficiently greater than 1.24 such that the decrease in calculated core l

ONBR after the trip will not result in a violation of the ONBR Safety Limit.

CPC uncertainties related to ONBR 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' i

time delays.

The ONBR algoritha 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.

Parameter Limitino Value a.

RCS Cold Leg Temperature-Low

> 470'F b.

RCS Cold Leg Temperature-High 7 610*F c.

Axial Shape Index-Positive Rot more positive than + 0.5 i

d.

Axial Shape Index-Negative Not more negative than - 0.5 e.

Pressurizer Pressure-Low

> 1860 psia l

f.

Pressurizer Pressure-High 7 2388 psia g.

Integrated Radial Peaking Factor-Low

> 1.28 h.

Integrated Radial Peaking Factor-High

< 4.28 i.

Quality Margin-Low IO Steam Generator Level - High The Steam Generator Level - High trip is provided to protect the turbine from excessive noisture 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 i:,olation setpoint.

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

i' PALO VERDE - UNIT 2 8 2-6 AMENOMENT NO. 19 4

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FIGURE 3.1-IA SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE PALO VERDE - UNIT 2 3/4 1-2a AMENDMENT No.19

FIGURE 3.1-1 l

ALLOWABLE NTC MODES 1 AND 2

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PALO VERDE UNIT 1 CYCLE 2 g

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00RF POWER LEVEL,I 0F RATED THERMAL POWER

,1 MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature (Tcold) shall be greater than or equal to 552*F.

APPLICABILITY:

MODES 1 and 2#*.

ACTION:

With a Reactor Coolant System operating locs temperature (Teold) less than 552*F, restore T to within its limit',ithin 15 minutes or be in HOT cold STANOBY within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tcold) shall be determined to be greater than or equal to 552*F:

a.

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

At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T,jq is less than 557'F.

c l

i

With K greater than or equal to 1.0.

g

See Special Test Exception 3.10.5.

PALO VERDE - UNIT 2 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.

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 3.1.1.2 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in l

at least HOT STANDRY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

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 .

c.

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 l

CEA(s) is either:

1.

Restored to OPERABLE status within its above specified : alignment j

requirements, or 2.

Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.2 is satisfied.

After declaring the CEA(s) l 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:

l 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 in-ches of the inoperable CEA(s) while maintaining the allow-able CEA sequence and insertion limits shown on Figures 3.1-3 and 3.1-4; the THERMAL POWER level shall be restricted pursuant to Specifications 3.1.3.6 and 3.1.3.7 during sub-

\\

sequent operation.

See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE - UNIT 2 3/4 1-21 AMENDMENT NO.19

REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.2 l

15 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.

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.

e.

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 REQUIREMENTS 4.1.3.1.1 The position of each full-length and part-length CEA shall be deter-mined 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 insertad 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 2 3/4 1-22 AMEN 0 MENT NO.19 l

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PALO VERDE - UNIT 2 3/4 1-23 AMENDMENT NO. 19

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TIME AFTER DEVIATION. MINUTES

WHEN CORE POWER IS REDUCED TO 55% OF RATED THERMAL 1

POWER PER THIS LIMIT CURVE. FURTHER REDUCTION IS NOT REQUIRED i

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FIGURE 3.1-2A a

CORE POWER LIMIT AFTER CEA DEVIATION" l

j PALO VERDE - UNIT 2 3/4 1-24 AMEN 0 MENT NO. 19 l

1

~

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS - OPERATING l

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:

With 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:

a.

Restore the inoperable position indicator channel to OPERABLE status, or b.

Be in at least HOT STANDBY, or c.

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.

Operation may then continue provided the CEA group (s) with the inoperable position indicator (s) is maintained fully withdrawn, except during surveillance 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 REQUIREMENTS 4.1.3.2 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.

4 PALO VERDE - UNIT 2 3/4 1-25 AMENDMENT NO. 19

~

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS - SHUT 00WN 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*, 48, and 5*.

ACTION:

With less than the above required position indicator channel (s) OPERA 8LE, immediately open the reactor trip breakers.

SURVEILLANCE REQUIREMENTS 4.1.3.3 The above required CEA Reed Switch Position Transmitter indicator channel (s) shall be determined to be OPERABLE by perfonnance of a CHANHEL FUNCTIONAL TEST at least once per 18 months.

"With the reactor trip breakers in the closed position.

I PALO VERDE - UNIT 2 3/4 1-26

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.PALO VERDE - UNIT 2 3/4 1 32 AMENOMENT NO.19 i

e 4-

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.

1 7 EFPD per 30 EFPD-interval, and b.

1 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 4.1.3.1.2, 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 EFP0 interval or > 14 EFPD per calendar year, either:

1.

Restore the part length group within the Long Term Steady State Insertion t.imit 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 REQUIREMENTS 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 .

1 j

See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE - UNIT 2 3/4 1-33 AMENDMENT NO. 19 i

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J 3/4.2 POWER OISTRIBUTION 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 within its limit using any operable CPC channel (when COLSS is out of service).

APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.

ACTION:

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

1.

COLSS calculated core pcwer 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:

l 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

.a.

b.

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

SURVEILLANCE REQUIREMENTS

~

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 l

THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring i

the core power distribution with the Core Operating Limit Supervisory System l

(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 withis, 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.

l

]

PALO VERDE - UNIT 2 3/4 2-1 AMENOMENT NO. 19

POWER DISTRIBUTION LIMITS 3/4.2.2 PLANAR RADIAL PEAKIE FACTORS - Fxy LIMITIE CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FAC 0RS (F" shall be less than or equaltcthePLANARRADIALPEAKINGFACTORS(F{@e)use Limit Supervisory System (COLSS) and in the C Protection Csiculators (CPC).

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER *.

ACTION:

c With an F"y exceeding a corresponding F y, 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 appliedtoplanarga glpeakingbyafactorequivalenttogreater than or equal to F andrestrictsubsequentopegaignsothata margin to the COLS5 o Nating limits of at least [(F, F,Y) - 1.0) x 100% is maintained; or b.

Adjust the affected PLANAR RADIAL PEAKIE FACTORS (FC ) used in the COLSS and CPC to a value greater,than or equal to th measured PLANAR RADIAL PEAK!E FACTORS (Fxy) or c.

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

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 The seasureri PLANAR RADIAL PEAKIE FACTORS (F" obtained by using theincoredetectionsystem,shallbgdeterminedtobeTe)ssthanorequalto the PLANAR RADIAL PEAKING FACTORS (Fxy), used in the COLSS and CPC at the following intervals:

a.

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

At least,once per 31 Effective Full Power Days.

See Special Test Excertion 3.10.2.

PALO VERDE - UNIT 2 3/4 2-2

y/

POWER DISTRIBUTION LIMITS 3/4.2.4 DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The ONBR 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 l

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 COL 35 is out of service and l

neither CEAC is operable).

APPLICABILITY:

MODE 1 above 20% of RATED THERMAL POWER.

ACTION:

With the ONBR not being inaintained:

1.

As indicated by COLSS calculated core power exceeding the appropriate COLSS calculated power operating limit; or 2.

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 inciease the ONBR 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 20% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 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 20% 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 DNBR, as indicated on any OPERABLE ONBR 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 l

operating limit based on DNBR.

PALO VERDE - UNIT 2 3/4 2-5 rMENOMENT NO. 19 j

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%D a

e a

COLSS ONBR POWER OPERATING LIMIT REDUCTIObl r/. OF RATED THERMAL POWER)

FIGURE 3.2-1 COL 55 DNBR POWER OPERATING LIMIT ALLOWANCE FOR BOTH CEACs INOPERABLE PALO VERDE - UNIT 2 3/4 2-6 AMEN 0 MENT N0. 19

44 COLSS OUT OF S(RVICE ONBR LIMIT LINE I

2.1 i

i i

ACCEPTABLE 2.0

~

OPERATION MINIMUM 1 CEAC OPERABLE 1.9 g

L1,1.85)

L2,1.85)

E 1.8

(.2.1.75) 1.7 UNACCEPTABLE OPERATION 1.6 I

i 1

'l 1

),$

-0.3

~-0.2

-0.1 0.0 9.1 9.2 0.3 CORE AVERAGE ASI FIGURE 3.2-2 DNBR MARGIN OPERATING LIMIT 8ASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs CPERA8LE)

(

PAL 0 VERDE - UNIT 2 3/4 2-7 AMENCHENT N0. 19

COLSS OUT OF SERVICE ONBR LIMIT Ll' E' N

2.4 ACCEPTABLE OPERATION 2.3 CEACs INOPERABLE

(.05.2.30)

(.2.2.30) g o

2.2 E

I

(.2.2.13) 2.1 Eu UNACCEPTABLE OPERATION 1

2.0 i

1.9

- 0.3

-0.2

- 0.1 0.0 0.1 0.2 0.3 i

CORE AVERADE ASI FIGURE 3.2-2A DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR $

~

(COLSS OUT OF SERVICE, CEACs INOPERA8LE)

PALO VERDE - UNIT 2 3/4 2-7a AMEN 0 MENT NO 19

POWER DISTRIBUTION LIMITS 3/4.2.5 RCS FLOW RATE 1

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 108 lbm/hr.

l 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 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 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 .

e 9

l f

PALO VERDE - UNIT 2 3/4 2-8 AMENDMENT NO.19

i g

POWER DISTRIBUTION LIMITS 3/4'. 2. 7 AXIAL SHAPE INDEX LIMITING CON 0! TION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:

a.

COLSS OPERABLE

-0.28 $ ASI 1 0.28 b.

COLSS OUT OF SERVICE (CPC).

-0.20 1 ASI i + 0.20 APPLICABILITY:

MODE 1 above 20% 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 20% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.7 The core average AXIAL SHAPi 1NDEX shall be determined to be within its limit at least once per 12 hotars using the COLS$ or any OPERA 8LE Core Protection Calculator channel.

1 "See Special Test Exception 3.10.2.

PALO VERDE - UNIT 2 3/4 2-11

a A.

s.

4 2

- +

4 POWER DISTRIBUTION LIMITS 1

t i

3/4.2.8 PRESSURIZER PRESSURE d

]

LIMITING CON 0! TION FOR OPERATION I

3.2.8 The pressurizer pressure shall be maintained between 2025 psia and 2300 psia.

I APPLICA8ILITY:

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 STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVE!LLANCE REQUIREMENTS i

4.2.8 The pressurizer pressure shall be determined to be within its limit at l

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

l l

l t

i i

1

)

See Special Test Exception 3.10.5

,i.

4 i

1 i

PALO VERDE - UNIT 2 3/4 2-12 AMENDMENT NO. 19 i

1 i

i I

TABLE 3.3-1 (Continued) j i

REACTOR PROTECTIVE INSTRUMENTATION ACTION STATEMENTS 3.

Steam Generator Pressure -

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

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

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

(Wide Range)

Steam Generator Level 1-Low (ESF)

Steam Generator Level.2-Low (ESF) 5.

Core Protection Calculator Local Power Density - High (RPS)

ON8R - Low (RPS)

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

Subsequent i

STARTUP and/or POWER OPERATION may continue if one channel is i

restored to OPERABLE status and the provisions of ACTION 2 are 1

satisfied.

l.

ACTION 4 With the number'of channels OPERA 8LE one less than required by i

the Minimus Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.

ACTION 5 With the number of channels OPERABLE one less than required j

by the Minimum Channels OPERA 8LE requirement, STARTUP and/or 1

POWER OPERATION may continue provided the reactor trip breaker l

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 STAN08Y within 6 3

hours; 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.

l l

l ACTION 6 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.

Wtt'. both CEACs inoperable, operation may continue provided l

4

[

that:

3 1.

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the ONBR 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 i

System is disabled, and I

PALO VERDE - UNIT 2 3/4 3-7 AMENOMENT NO 19 1

1

s/

I TABLE 3.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION 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, b)

The "RSPT/CEAC Inoperable" addressable con' tant s

in the CPCs is set to indicate that both CEACs are inoperable, 4

c)

The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently 1

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 full-length and part-length CEAs are verified fully withdrawn except j

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 j

per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the inserted CEAs are aligned within j

6.6 inches (indicated position) of all other CEAs in its group.

ACTION 7 With three or more auto restarts, excluding periodic auto l

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 .

ACTION 8 With the number of OPERA 8LE channels one less than the Minimum Channels OPERA 8LE requirement, restore an inoperable channel to OPERA 8LE 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 2 3/4 3 8 AMENOMENT NO. 19

e THIS PAGE INTENTIONALLY LEFT BLANK.

g.

PALO VERDE - UNIT 2 3/4 3-9 AMENOMENT NO. 19

l l

l i

l l

1 1

1 TilIS PAGE INTENTIONALLY LEIT Bl.ANK, j

1 i

i l

l l

PALO VERDE - UNIT 2 3/4 3-10 AMENOMENT NO, 19

TABLE 3.3-2 E

o REACTOR PROTECTIVE INSTRtMENTATION RESPONSE TIMES Mg FUNCTIONAL UNIT RESPONSE TIME 7

1.

TRIP GENERATION g

A.

Process U

1.

Pressurizer Pressure - High

$ 1.15 seconds u

2.

Pressurizer Pressure - Low 1 1.15 seconds 3.

Steam Generator Level - Low

$ 1.15 seconds 4.

Steam Generator Level - High

$ 1.15 seconds

)

5.

Steam Generator Pressure - Low

$ 1.15 seconds 6.

Containment Pressure - High

< 1.15 seconds 7.

Reactor Coolant Flow - Low

$ 0.58 second w

z.

8.

Local Power Density - High ok a.

Net: tron Flux Power from Excore Neutron Detectors

< 0.75 second*

b.

CEA Positions i 1.35 second**

c.

CEA Positions:

CEAC Penalty Factor 30.75second**

9.

DNBR - Low a.

Neutron Flux Power from Excore Neutron Detectors

< 0.75 seconda b.

CEA Positions

< 1.35 second**

c.

Cold Leg Temperature 7 0.75 secondf#

t d.

Hot Leg Temperature E 0.75 secondff e.

Primary Coolant Pump Shatt Speed 2 0.30 secondt l

f.

Reactor Coolant Pressure from Pressurizer E 0.75 secondf##

g g.

CEA Positions:

CEAC Penalty Factor

30.75second**

l B.

Excore Neutron Flux 3

1.

Variable Overpower Trip

$ 0.55 second*

g 2.

Logarithmic Power Level - High a.

Startup and Operating

< 0.55 second*

U b.

Shutdown

{0.55second*

TABLE 3.3-2 (Continued) y REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT 7

C Core Protection Calculator System g

1.

CEA Calculators Not Applicable U

2.

Core Protection Calculators Not Applicable D.

Supplementary Protection System Pressurizer Pressure - High

< 1.15 second

11. RPS LOGIC A.

h trix Logic No* Applicable B.

Initiation Logic Not Applicable III.

RPS ACTUATION DEVICES

{

A.

Reactor Trip Breakers Not Applicable 8.

k nual Trip Not Applicable M

~

Neutron detectors are exempt from res m se time testing.

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

At 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.

The pulse transmitters measuring pump sprad are except from response time testint).

The k

response time shall be measured from the pulse shaper input.

6

1. Response time shall be measured from the output of the resistance temperature detector A

(sensor).

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

The measured 4

response time of the slowest RID shall be less than or equal to 8 seconds.

8

1. 1. Response time shall be measured from the output cf the pressure transmitter.

The

(

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

- - ~

b 9

i o

e THIS PAGE INTENTIONALLY LEFT BLANK.

f e

0 PALO VERDE - UNIT 2 3/4 3-13 AMEN 0 MENT NO. 19

i 4

TABLE 4.3-1 5

REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-o l

M E

CHANNEL MODES IN WHICH 1

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

[

FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED h

I.

TRIP GENERATION N

A.

Process 1.

Pressurizer Pressure - High 5

R M

1, 2 2.

Pressurizer Pressure - Low S

R M

1, 2 3.

Steam Generator Level - Low S

R M

1, 2 4.

Steam Generator Level - High 5

R H

1, 2 I

5.

Steam Generator Pressure - Low S

R M

1, 2, 3*, 4*

6.

Containment Pressure - High S

R H

1, 2 y

7.

Reactor Coolant Flow - Low S

R M

1, 2 8.

Local Power Density - High S

0 (2, 4), R (4, 5)

M, R (6) 1, 2 9.

DNBR - Low S

0 (2, 4), R (4, 5)

M, R (6) 1, 2 M (8), S (7)

I 8.

Excore Neutron Flux 1.

Variable Overpower Trip 5

0 (2, 4), M (3, 4)

M 1, 2 Q (4) 2.

Logarithmic Power Level - High S

R (4)

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

C.

Core Protection Calculator System 1.

CEA Calculators S

R M, R (6) 1, 2 2.

Core Protection Calculators S

0 (2, 4), R (4, 5)

M (9), R (6) 1, 2 M (8), S (7) e G

e a

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM It$TRUMENTATION TRIP VALUES q

E ESFA SYSTEM FUNCTIONAL UNIT TRIP SETPOIN1 ALLOWABLE VALUES 7

I.

SAFETY INJECTION (SIAS) g A.

Sensor / Trip Units M

1.

Containment Pressure - High

$ 3.0 psig

$ 3.2 psig 2.

Pressurizer' Pressure - Low 1 1837 psia (1) 1822 psia (I) 8.

ESFA System Logic Not Applicable Not Applicable C.

Actuation Systems Not Applicable Not Applicable II.

CONTAI MENT ISOLATION (CIAS)

A.

Sensor / Trip Units 1.

Containment Pressure - High 5 3.0 psig

$ 3.2 psig

{

2.

Pressurizer Pressure - Low 1 1837 psia 1 1822 psia (1)

II) 8.

ESFA System Logic Not Applicable Not Applicable i

y j

UI C.

Actuation Systems Not Applicable Not Applicable III. CONTAIMENT SPRAY (CSAS)

)

A.

Sensor / Trip Units j

Containment Pressure High - High

$ 8.5 psig

$ 8.9 psig 1

B.

ESFA System Logic Not Applicable Not Applicable i

C.

Actuation Systems Not* Applicable Not Applicable f

IV.

MAIN STEAM LINE ISOLATION (NSIS)

I A.

Sensor / Trip Units 1.

Steam Generator Pressure - Low 1 919 psia (3) 912 psia (3)

I2) l 2.

Steam Generator Level - High 5 91.0% NR(2)

$ 91.5% NR

}

3.

Containment Pressure - High

$ 3.0 psig 5 3.2 psig 8.

ESFA System Logic Not Applicable Not Applicable C.

Actuation Systems Not Applicable Not Applicable I

l s

-s-

TABLE 3.3-4 (Continued)

,>6 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES 4

ESFA SYSTEM FUNCTIONAL UNIT TRIP VALUES ALLOWABLE VALUES 7

V.

RECIRCULATION (RAS)

A.

Sensor / Trip Units g

Z Refueling Water Storage Tank - Low 7.4% of Span 7.9 1 % of Span 1 6.9 8.

ESFA System Logic Not Applicable Not Applicable C.

Actuation Syste's Not Applicable Not Applicable VI. AUXILIARYFEEDWATER(SG-1)(AFAS-1)

A.

Sensor / Trip Units I4)

I4) 1.

Steam Generator #1 l.evel - Low 1 25.8% WR 1 25.3% WR 1

SG2 > SG1

-< 185 psid

-< 192 psid 2.

Steam Generator a Pressure -

w B.

ESFA System Logic Not Applicable Not Applicable C.

Actuation Systems Not Applicable Not Applicable VII. AUXILIARY FEEDWATER (SG-2)(AFAS-2)

A.

Sensor / Trip Units f4) 1 25.3% WR(4) 1.

Steam Generator #2 Level - Low

> 25.8% WR 2.

Steam Generator A Pressure -

1 185 psid i 192 psid SG1 > SG2 B.

ESFA System Logic Not Applicable Not Applicable g

C.

Actuation Systems Not Applicable Not Applicable E

VIII.

LOSS OF POWER E

A.

4.16 kV Emergency Bus Undervoltage

'i (Loss of Voltage)

> 3250 volts

> 3250 volts s

B.

4.16 kV Emergency Bus Undervoltage 2930 to 3744 volts 2930 to 3744 volts G

with a 35-second with a 35-second (Degraded Voltage) maximum time delay maximum time delay IX.

CONTROL ROOM ESSENTIAL FILTRATION-1 2 x 10 5 pCi/cc 1 2 x 10 5 pCi/cc

\\

~

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN ANC K

- CEA WORTH TESTS g

LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN *and K requirements of Specification 3.1.1.2 may l

u3 be suspended for measurement of CE worth and shutdown margin provided reac-tivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s), or the reactor is suberitical by at least the reactivity equivalent of the highest CEA worth.

APPLICABILITY:

MODES 2, 3* and 4*#.

ACTION:

a.

With any full-length CEA-not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN and K required by Specifica-N-1 tion 3.1.1.2 are restored.

b.

With all full-length CEAs fully inserted and the reactor suberitical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 26 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REOUIREMENTS 4.10.1.1 'The position of each full-length and part-length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

i 4.10.1.3 When in MODE 3 or MODE 4, the reactor shall be determined to be suberitical by at least the reactivity equivalent of the highest estimated CEA worth or the reactivity equivalent of the highest estimated CEA worth is avail-able for trip insertion from OPERABLE CEAs at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by con-sideration of at least the following factors:

Reactor _ Coolant System boron concentration, a.

b.

CEA positio'",

n c.

Reactor Coolant System average temperature, d.

Fuel burnup based on gross thermal energy generation.

T-e.

Xenon concentration, and f.

Samarium concentration.

0peration in MODE 3 and MODE 4 shall be limited to 6 consecutive hours.

Limited to low power PHYSICS TESTING at the 320*F plateau.

PALO VERDE - UNIT 2 3/4 10-1 AMENOMENT NO.13

SPECIAL TEST EXCEPTIONS 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION l

3.10.2 The moderator temperature coefficient, group height, insertion, and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and b.

The limitt of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODES 1 and 2.

j ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.l.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.

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

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended and shall be verified,_to be within the test power plateau.

4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, l

3.2.2, 3.2.3, 3.2.7, or the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended.

PALO VERDE - UNIT 2 3/4 10-2 AMENDMENT NO. 19

o SPECIAL TEST EXCEPTIONS 3/4.10.3 REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specification 3.4.1.1 and noted requirements of Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup PHYSICS TESTS, provided:

a.

The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b.

The reactor trip setpoints of the OPERABLE power level channels are set at less than or equal to 20% of RATED THERMAL POWER.

~

c.

Both reactor coolant loops and at least one reactor coolant pump in each loop are in operation.

I APPLICABILITY:

Ouring startup PHYSICS TESTS.

ACTION:

l With the THERMAL POWER greater than 5% of RATE 0' THERMAL POWER or with less than the above required reactor coolant loops in operation and circulating reactor coolant, immediately trip the reactor.

SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%

of RATED THERMAL POWER at least once per hour during startup PHYSICS TESTS. ;

4.10.3.2 Each logarithmic and variable overpower level neutron flux monitoring channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating startup PHYSICS TESTS.

4.10.3.3 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I 6

PALO VERDE - UNIT 2 3/4 10-3

-.m..i.-----

--.-.-,w, 4

u

--.y

-~ -- --

SPECIAL TEST EXCEPTIONS 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLO LEG TEMPERATURE LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.1.3.1, 3.1.3.6, 3.1.3.7, and 3.2.6 may be suspended during the performance of PHYSICS TESTS to determine the isothermal temperature coefficient, moderator temperature coefficient, and power coefficient provided the limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.4.2 below.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.6, 3.1.3.7, and 3.2.6 are j

suspended, either:

a.

Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or b.

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

SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.1, 3.1.3.6, 3.1.3.7, and/or 3.2.6 are suspended and shall be verified to be within the test power plateau.

4.10.4.2 The linear heat rate shall be determined to be within tha limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specification 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1, 3.1.3.6, 3.1.3.7, and/or 3.2.6 are suspended.

PALO VERDE - UNIT 2 3/4 10-4 AMENDMENT NO.19

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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 Tcold greater than or 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 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 o'f 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 VER0E - UNIT 2 B 3/4 1-5

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@CTIVITYCONTROLSYSTEMS 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 consistert set of radial peaking factors defined.

The Long Term Steady State and Short Term Insertion Limits are deter-mined based upon the assumeo 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 behav.ior of the radial peaking factors within the bounds determined from analysis.

The actions specified serve to limit 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 l

specified for the plant which has been designed for primarily base loaded operation 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 the l

Shutdown CEA Insertson Limits of Specification 3.1.3.5 ensure that (1) the minimum SHUT 00WN 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 ONB Power Operating Limit (POL) and various operating parameters 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 (A00), 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 l

Service (C005), the monitoring ~ function is performed via the CPC calculation of 1

DNBR in conjunction with Technical Specification C005 Limit Lines (Figures 3.2-2 and 3.2-2A) which restricts the reactor power sufficiently to preserve the R0PM.

I 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 ONB 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 2 B 3/4 1-6 AMEN 0 MENT NO. 19

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 R0PM 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 6:cordance with Figure 3.1-5 to justify reduction of the PLR deviation penalty l

factors.

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

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PALO VERDE - UNIT 2 B 3/4 1-7 AMENDMENT NO. 19

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 Systen (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 l

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 factors 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 ONB, 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 2 8 3/4 2-1 AMENDMENT NO. 19

POWER OISTRIBUTION LIMITS BASES 3/4.2.2 PLANAR RADIAL PEAKING FACTORS LimitingthevaluesofthePLANARRADIALPEAKINGFACTORS(Fy)usedinthe COLSS and CPCs to values equal to or greater than the measured PLANAR RADIAL PEAKING FACTORS (Fxy) provides assurance that the limits calculated by COLSS and the CPCs remain valid.

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

A minimum core power l

at 20% of RATEC THERMAL POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS.

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

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

The periodic l

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 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

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

An AZIMUTHAL POWER TILT greater than j

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 ti'it is normally calculated by COLSS.

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

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

Core nois,e 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 (P

/Puntilt)-1.0 where:

tilt 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:

P

/P

=1+T g cos (0 - Oo) tilt untilt q

where:

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

g is the radial normalizing factor 8 is the azimuthal core location i

Oo is the azimuthal core location of maximum tilt PALO VERDE - UNIT 2 8 3/4 2-2

POWER DISTRIBUTION LIMITS BASES AZIMUTHAL POWER TILT - Tq (Continued)

P

/P is the ratio of the power at a core location in the presence tilt untilt 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 DN8R MARGIN The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conserv3tive 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 minie.um 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 l

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) is 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, sof tware algorithm modelling, computer processing, rod bow, and core power measurement.

Parameters required to maintain the margin to DNB and total core power i

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 l

maintained by utilizing a predetermined DN8R 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 20% of RATED THERMAL POWER.

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

l Core noise level at lo,w power is too large to obtain usable detector readings.

A DNBR penalty factor has been included in the COLSS and CPC DNBR calcula-tions 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 compensate 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 2 8 3/4 2-3 AMENDMENT NO.19

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POWER DISTRIBUTION LIMITS BASES 3/4.2.5 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 95% of the design flow rate (164.0 x 108 lbm/hr) or 155.8 x 106 lbm/hr.

The actual RCS flow rate is deter-mined by direct measurement and an uncertainty associated with that measurement is considered when comparing actual RCS flow rate to the minimum required value of 155.8 x 108 lbm/hr.

3/4.2.6 REACTOR COOLANT COLO 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 2 B 3/4 2-4 AMEN 0 MENT NO. 19

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

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

The integrated operation of each of these systems is consistent with the assumptions i

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.

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.

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

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

PALO VERDE - UNIT 2 B 3/4 3-1 AMENOMENT NO. 19

INSTRUMENTATION BASES l

REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENiATION (Continued)

(

The measurement of response time at the specified frequencies provides l

assurance that the protective and ESF action function associated with each l

channel is completed within the time limit assumed in the safety analyses.

I No credit was taken in the analyses for those channels with response times f

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 i

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

Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either (1) in place, onsite, or offsite test measurements or (2) utilizing replacement sensors with certified response times.

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1 PALO VERDE - UNIT 2 B 3/4 3-2 AMEN 0 MENT NO. 19

d INSTROMENTATION BASES 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that:

(1) the radiation levels are continually measured in the areas served by the individual channels and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 INCORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum comple-ment of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

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

The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, "Instrumentation for Earth-quakes," April 1974 as identified in the PVNGS FSAR.

The seismic instrumenta-tion for the site is listed in Table 3.3-7.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that suffi-cient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February 1972.

Wind speeds less than 0.6 MPH cannot be measured by the meteorological instrumentation.

3/4.3.3.5 REMOTE SHUTOOWN SYSTEM The OPERABILITY of the remote shutdown system ensures that sufficient capability is available to permit safe shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

PALO VERDE - UNIT 2 8 3/4 3-3 AMENOMENT NO. 19

INSTRUMENTATION BASES REMOTE SHUTDOWN SYSTEM (Continued)

The parameters selected to be monitored ensure that (1) the condition of the reactor is known, (2) conditions in the RCS are known, (3) the steam generators are available for residual heat removal, (4) a source of water is available for makeup to the RCS, and (5) the charging system is available.to makeup water to the RCS.

The OPERABILITY of the remote shutdown system insures that a fire will not preclude achieving safe shutdown.

The remote shutdown system instrumenta-tion, control and power circuits and disconnect switches necessary to eliminate effects of the fire and allow operation of instrumentation, control and power circuits required to achieve and maintain a safe shutdown condition are independent of areas where a. fire could damage systems normally used to shutdown the reactor.

This capability is consistent with General Design Criterion 3 and Appendix R to 10 CFR 50.

The alternate disconnect methods or power or control circuits ensure that sufficient capability is available to permit shutdown and maintenance of cold shutdown of the facility by relying on additional operator actions at local control stations rather than at the RSP.

l 3/4.3.3.6 POST-ACCIDENT MONITORING INSTRUMENTATION l

The OPERABILITY of the post-accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.

This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

The containment high range area monitors (RU-148 & RU-149) and the main steamline radiation monitors (RU-139 A&8 and RU-140 A&B) are in Table 3.3-6.

The high range effluent monitors and samplers (RU-142, RU-144 and RU-146) are in Table 3.3-13.

The containment hydrogen monitors are in Specifica-I tion 3/4.6.4.1.

The Post Accident Sampling System (RCS coolant) is in Table 3.3-6.

l The Subcooled Margin Monitor (Sle), the Heat Junction Thermocouple (HJTC),

and the Core Exit Thermocouples (CET) comprise the Inadequate Core Cooling (ICC) instrumentation required by Item II.F.2 NUREG-0737, the Post TMI-2 Action Plan.

The function of the ICC instrumentation is to enhance the ability of the plant operator to diagnose the approach to existance of, and recovery from ICC.

Additionally, they aid in tracking reactor coolant inventory.

These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37.

These are not required by the accident analysis, nor to bring the plant to Cold Shutdown.

PALO VERDE - UNIT 2 B 3/4 3-4 AMENOMENT NO.19 1

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INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT MONITORING INSTRUMENTATION (Continued)

In the event more than four sensors in a Reactor Vessel Level channel are inoperable, repairs may only be possible during the next refueling outage.

This is because the sensors are accessible only after the missile shield and reactor vessel head are removed.

It is not feasible to repair a channel except during a refueling outage when the missile shield and reactor vessel head are removed to refuel the core.

If both channels are inoperable,.the channels shall be restored to OPERABLE status in the nearest refueling out-age.

If only one channel is inoperable, it is intented that this channel be restored to OPERABLE status in a refualing outage as soon as reasonably possible.

3/4.3.3.7 LOOSE-PART DETECTION INSTRUMENTATION The OPERABILITY of the loose part detection instrumentation ensures that sufficient capability is available to detect loose metallic parts in the primary systern and avoid or mitigate damage to primary system components.

The allowable outm f-service times and surveillance requirements are consistent with the recor<cendations of Regulatory Guide 1.133, "Loose-Part Detection Program for the F+imary System of Light-Water-Cooled Reactors," May 1981.

3/4.3 g RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Ti* radioactive gaseous effluent instrumentation is provided to monitor and control, is applicable, the releases of radioa-tive materials in gaseous effluents during actual or potential releases of gaseous effluents.

The alarm / trip set-points for these instruments shall be calculated and adjusted in accordance with the methodology and paramete.rs in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20.

This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the GASEOUS RADWASTE SYSTEM.

The OPERA-BILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

There are two separate radioactive gaseous effluent monitoring systems:

the low range effluent monitors for normal plant radioactive gaseous effluents and the high range effluent monitors for post-accident plant radioactive gaseous effluents.

The low range monitors operate at all times until the concentration of radioactivity in th'e effluent becomes too high during post-accident conditions.

The high range monitors only operate when the concentration of radioactivity in the effluent is above the setpoint in the low range monitors.

I PALO VERDE - UNIT 2 8 3/4 3-5 AMEN 0 MENT NO. 19 L

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