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Corrected Ltr Forwarding Amend 18 to License NPF-74 & Safety Evaluation.Amend Incorporates Cycle 2 Changes in Tech Specs
	ML20245J224
Person / Time
Site:	Palo Verde
Issue date:	06/23/1989
From:	Marlone Davis; Office of Nuclear Reactor Regulation
To:	Conway W; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
References
	TAC-71574, NUDOCS 8906300244
	Download: ML20245J224 (3)
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RE-DISTRIBUTED 6/23/89 - l June 9, 1989 j

j Docket No.: STN 50-530 i

Mr. William F. Conway l Executive Vice President Arizona Nuclear Power Project f Post Office Box 52034 t i

Phoenix, Arizona 85072-2034

Dear Mr. Conway:

SUBJECT:

ISSUANCE OF AMEN 0 MENT NO.18 TO FACILITY OPERATING LICENSE NO NPF-74 FOR THE PALO VERDE NUCLEAR GENERATING STATION, UNIT 3 (TAC NO. 71574)

) 1 The Comission has issued the subject Amendment, which is enclosed, to the t Facility Operating License for Palo Verde Nuclear Generating Station, Unit 3. 1 The Amendment consists of a change to the Technical Specifications (Appendix A to the license) in response to your application transmitted by letter dated December 14, 1988. I The Amendment revises several portions of the Technical Specifications to incorporate changes in support of Cycle 2 operation for Palo Verde, Unit 3. g A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Regis,ter notice.

. Sincerely, P 000ll g00 O623 o

Mic ael J. Davis, Project Manager Project Directorate V Division of Reactors Projects - III, IV, Y and Special Projects

Enclosures:

1. Amendment No. 18 to NPF-74 '

2. Safety Evaluation I

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Mr. William F. Conway Executive Vice President Arizona Nuclear Power Project Post Office Box 52034 Phoenix, Arizona 85072-2034

Dear Mr. Conway:

SUBJECT:

ISSUANCE OF AMENDMENT NO. 18 TO FACILITY OPERATING LICENSE NO. NPF-74 FOR THE PALO VERDE NUCLEAR GENERATING STATION, UNIT 3 (TAC NO. 71574)

The Cossnission has issued the subject Amendment, which is enclosed, to the Facility Operating License for Palo Verde Nuclear Generating Station, Unit 3.

The Amendment consists of a change to the Technical Specifications (Appendix A to the license) in response to your application transmitted by letter dated December 14, 1988.

The Amendment revises several portions of the Technical Specifications to incorporate changes in support of Cycle 2 operation for Palo Verde, Unit 3.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Comission's next regular biweekly Federal Register notice.

Sincerely, r,

,ffw L ly 2.w'o

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i Michael J. Davis, Project Manager Project Directorate V Division of Reactors Projects - III, IV, V and Special Projects

Enclosures:

1. Amendment No. 18 to NPF-74

2. Safety Evaluation j cc: See next page y h

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Mr. William F. Conway Palo Verde Arizona Nuclear Power Project Cc:

Mr. William F. Conway Arizona Nuclear Power Project-Executive Vice President Post Office' Box 52034 Phoenix, Arizona 85072-2034 Arthur C. Gehr, Esq.

Snell. & Wilmer 3100 Valley Center Phoenix, Arizona 85073 Charles.R. Kocher, Esq. Assistant Council James A. Boeletto, Esq.

Southern California Edison Company P. O. Box 800 Rosemead, California 91770 Mr. Tim Polich U.S. Nuclear Regulatory Commission HC-03 Box ~293-NR j.

Buckeye, Arizona 85326 l

Regional Administrator, Region V U. S. Nuclear Regulatory Commission i 1450 Maria Lane !

Suite 210 Walnut Creek, California 94596 Mr. Charles B. Brinkman Washington Nuclear Operations 1 Combustion Engineering, Inc.

12300 Twinbrook Parkway, Suite 330 '

Rockville, Maryland 20852 Mr. Charles Tedford, Director Arizona Radia . Regulatory Agency 4814 South 40 t Phoenix, Ari , 85040 Chairman Maricopa County Board of Supervisors 111 South Third Avenue l Phoenix, Arizona 85003 1

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ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. 50-530 PALO VERDE NUCLEAR GENERATING STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 18 License No. NPF-74

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

A. The application for amendment, dated December 14, 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 Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority (licensees), complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C. Thereisreasonableassurance(i)thattheactivitiesauthorized 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 common i defense and security or to the health and safety of the public; E. The 1ssuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

l. 2. Accordingly, the license is amended by changes to the Technical L Specifications as indicated in the enclosure to this license amendment, i and paragraph 2.C(2) of Facility Operating License No. NPF- 74 is hereby amended to read as follows:

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l (2) Technical Specifications and Environment 1' Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.18, 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 an'd 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 REGULATORY COMMISSION d6 Y st George .-Knighton rector Project Directorate V Division of Reactor Projects III, IV, V and Special Projects

Enclosure:

Changes to the Technical Specifications Date of Issuance: June 9, 1989 t

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ENCLOSURE TO LICENSE AMENDMENT ;

AMENDMENT NO. 18 TO FACILITY OPERATING LICENSE NO. NPF-74 DOCKET NO. STN 50-530 Replace the following pages of the Appendix A Technical Specifications with ,

the enclosed pages. The revised pages are identified by Amendment number and I contain vertical lines indicating the areas of change. Also to be replaced. '

are the following overleaf pages to the amended pages, i Amendment Pages Overleaf Pages IV III ;

XIX -

XX -

l 2-1 2-2 \

2-3 2-4 2-5 -

B 2-1 -

B 2-2 -

B 2-3 B 2-4 !

B 2-5 -

B 2-6 -

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ENCLOSURE T0 LICENSE AMENDMENT AMENDNENT NO. 18 TO FACILITY OPERATING LICENSE NO. NPF-74 DOCKET NO. -STN 50-530' Replace-the'following 3 ages of the Appendix N Technical Specifications with

.the enclosed pages. . T1e revised pages are identified by Amendment number and contain-vertical lines indicating the. areas of change. Also to be replaced. .i are the following overleaf pages to the amended pages. '

Amendment Pages Overleaf Pages IV III XIX -

XX -

2-1 2-2 2-3 2-4 2-5 -

B 2-1 -

B-2-2 -

B 2-3 B 2-4 B.2-5 -

B 2-6 -

3/4 1-2a -

3/4 1-5 3/4 1-6 3/4 1-21 -

3/4 1 -

3/4 1-23 -

3/4 1-24 -

3/4 1-25 3/4 1-26 3/4 1-28 3/4 1-27 3/4 1-29 -

3/4 1-30 -

3/4 1-30A -

3/4 1-30s -

3/4 1- -

L 3/4 1- -

3/4 1- -

3/4 2-1 3/4 2-2 3/4 2-4a - l 3/4 2-5 -

l 3/4 2-6 - ;

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)

3/4 2-8 -

3/4 2-12 3/4 2-11 e

j; 9. .,7-4.

DOCKET NO. STN'50-530 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 i

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B 3/4 3-4 -

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INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l

l SECTION.

PAGE l

2.1 SAFETY LIMITS I

2.1.1 REACTOR C0RE.............. i 2.1.1.1 .............................. 2-1 0NBR...................... ............... .............. 2-1 l

2.1.1.2. PEAK-LINEAR HEAT RATE..................... 1 2.1.2 ........ ..... 2-1 REACTOR COOLANT SYSTEM PRESSURE.................... ..... 2-1 i 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SETP0lNTS.....................................

2-2 1

k 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 SETP0!NTS..................................... B 2-2 PALO VERDE - UNIT 3 III i

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l INDEX l

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' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

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SECTION PAGE j L3/4.0 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 - K N ANY CEA WITHDRAWN............ 3/4 1-2 MODERATOR TEMPERATURE COEFFICIENT..................... 3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY................... 3/4 1-6 3/4.1.2 B0 RATION SYSTEMS F LOW P ATH S - SHUTD0WN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-7 FLOW PATHS - 0PERATING................................ 3/4 1-8 CHARGING PUMPS - SHUTD0WN............................. 3/4 1-9 CHARGING PUMPS - 0PERATING............................ 3/4 1-10 BORATED WATER SOURCES - SHUTD0WN...................... 3/4 1-11 BORATED WATER SOURCES - 0PERATING..................... 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 - SHUTDOWN................ 3/4 1-26 CEA DROP 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 1-30A PALO VERDE - UNIT 3 IV AMENDMENT NO.18

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INDEX LIST OF FIGURES PAGE ;

3 1-1A SHUTDOWN MARGIN VERSUS COLD LEG TEMPERATURE............ 3/4 1-2a 3.1-1

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ALLOWABLE MTC MODES 1 AND 2.... 3/4 1-5 3.1-2 MINIMUM BORATED WATER V0LUMES................. ......... 3/4 1-12 3.1-2A PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER....... 3/4 1-23 3.1-2B CORE POWER LIMIT AFTER CEA DEVIATION................... 3/4 1-24 4 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 1-33 l 3.2-1A AZIMUTHAL POWER TILT LIMIT VS. THERMAL POWER (COLSS IN SERVICE)..................................... 3/4 2-4a 3.2-1 COLSS DNBR 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 PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs IN0PERABLE)... 3/4 2-7a 3.2-3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVEL.................................................. 3/4 2-10 1

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 B 3/4.4-1 N ' DUCTILITY TRANSITION TEMPERATURE INCREASE AS A F ION OF FAST (E > 1 MeV) NEUTRON FLUENCE (550*F IRRADIATION).................................... B 3/4 4-10 5.1-1 SITE AND EXCLUSION B0VNDARIES.......................... 5-2 5.1-2 LOW PO P U LAT I O N Z0 N E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 5.1-3 GASEOUS RELEASE P0INTS................................. 5-4 PALO VERDE - UNIT 3 XIX AMENDMENT NO. 18

INDEX LIST 0F TABLES PAGE 1.1 FREQUENCY N0TATION...................................... 1-8

1. 2 OPERATIONAL M00ES....................................... 1-9 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP S LIMITS...................................ETPOINT ............... 2-3 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL M00E$.....................

3.1- 1 FOR K,77 > 0.98......................................... 3/4 1-16 3.1-2 FOR 0.98 > K,7f > 0.97.................................. 3/4 1-17 3.1-3 FO R 0. 9 7 > K,7 f > 0. 9 6. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-18 3.1-4 FOR 0.96 > K,ff > 0.95..................................

_ 3/4 1-19 3.1-5 FOR K,77 5,0.95......................................... 3/4 1-20 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 1 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 ENGINEERED SAFETY FEATURES RESPONSE TIMES............... 3/4 3-28 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3 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 SEIS11C MONITORING INSTRUMENTATION...................... 3/4 3-43 i 4.3-4 SE M 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 l PALO VERDE - UNIT 3 XX AMENDMENT NO. 18

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.c 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR 2.1.1.1 The calculated DNBR of the reactor core shall be maintained greater than or equal.to.l.24.

l APPLICABILITY: MODES 1 and 2.

. ACTION:

Whenever the calculated DNBR of the reactor has decreased to less than 1.24, l 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.-

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 STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and comply with the requirements of Specification 6.7.

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. j J

ACTION:

MODES 1 and 2:

Wheneverthe$3pctorCoolantSystempressurehasexceeded2750 psia,beinHOT STANDBY with t m Reactor Coolant System pressure within its limit within 1 !

hour, and comply with the requirements of Specification 6.7.

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.

PALO VERDE - UNIT 3 2-1 AMENDMENT NO.18

SAFETY LIMITS AND LIMITING SAFsTY 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 limits 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 Setpoint value.

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. I I I I I nE q, E-lllll 1 l j

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS 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 j 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 pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.

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

(4) % 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 4 the trip setpoint includes measurement, calculational and processor uncer-tainties. Trip may be manually bypassed below 10 4% of RATED THERMAL l

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

(6) RATE is the maximum rate of decrease of the trip setpoint. There are no restrictions on the rate at which the setpoint can increase.

FLOLR is the minimum value of the trip setpoint.

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

Setpoints are based on steam generator differential pressure. 1 (7) The setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.

(8) RATE is the maximum rate of increase of the trip setpoint. (The rate at which the setpoint can decrease is no slower than five percent per second.)

CEILING is the maximum value of the trip setpoint.

BAND is the amount by which the trip setpoint is above the steady state input signal unless limited by the rate or the ceiling.

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

PALO VERDE - UNIT 3 2-5 AMENDMENT NO. 18

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 l of fission products to the reactor coolant. Overheating of the fuel cladding i is prevented by (1) restricting fuel operation to within the nucleate boiling j regime where the heat transfer coefficient is large and the cladding surface i temperature is slightly above the coolant saturation temperature, and n (2) maintaining the dynamically adjusted peak linear heat rate of the fuel !

at or less than 21 kW/ft which will not cause fuel centerline melting in any (

fuel rod.

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

The upper boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.

Correlations predict DNB and the location of DNB for axially uniform and I 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 l 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 whicW 4epends 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 at"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.

I PALO VERDE - UNIT 3 B 2-1 AMENDMENT NO.18

4 .

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

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

The Reactor Coolant System components are designed to Section III, 1974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 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. l 2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than ;

the drift allowance assumed for each trip in the safety analyses. '

The DNBR - Low and Local Power Density - High are digitally generated trip setpoints based on Safety Limits of 1.24 and 21 kW/ft, respectively. l Since these trips are digitally generated by the Core Protection Calculators, I the trip valuesJare not subject to drifts common to trips generated by analog type equipme $ ' The Allowable Values for these trips are therefore the same as the Trip Se$oints.

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 I 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." l l

}

PALO VERDE - UNIT 3 B 2-2 AMENDMENT NO. 18 l

_ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ _ _ _ -l

4 1

' BASES j

j 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 - R instrumentation channels and provides manual reactor, trip capability.

V.ariable 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 setpoints 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 System pressure boundary in the event of an'. unplanned criticality from a shutdown condition. A reactor trip is initiated by the Logarithmic Power Level - High trip unless this trip is manually bypassed by the operator. The operator may manually bypass this-trip when the THERMAL POWER level .is above 10 4% of RATED THERMAL POWER;.this bypass is automatically removed when the THERMAL POWER level decreases to 10 *% of RATED THERMAL POWER.

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

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

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

I 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 l PALO VERDE - UNIT 3 8 2-3 AMENDMENT NO. 18

BASES 1 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,lprovided the margin between the pressurizer pressure and this trip's setpoint is maintained.at less than or equal to 400 psi; this setpoint: increases automatically as:

pressurizer pressure, increases until the-trip setpoint is reached. The operator may manually bypass this trip when pressurizer pressure is below 400 psia. This bypass is' automatically removed when the-pressurizer pressure increases to 500 psia. _

Containment Pressure - High l

]

The Containment Pressure - High trip provides assurance that a reactor l I

trip is initiated in the event of containment building pressurization due to a '

pipe break inside the containment building. The setpoint for.this trip is identical to the safety injection setpoint. ,

Steam Generator Pressure - Low I The Steam Generator Pressure - Low trip provides protection in the event-of an increase.in heat removal by the secondary system and subsequent cooldown of.the reactor coolant. The setpoint is sufficiently below the full load operating point so as not to interfere with normal operation, but still high enough to provide the required protection in the event of. excessively high steam flow. This trip's setpoint may be manually decreased as steam generator pressure is reduced during plant shutdowns, provided the margin between the-steam generator. pressure and this trip's setpoint'is maintained at less than or equal-to 200 psi; this setpoint increases automatically as steam generator pressure increases until the normal pressure trip setpoint is reached.

Steam Generator Level - Low The Steam Generator Level - Low trip provides protection agai*nst a loss of feedwater flow incident and assures that the. design pressure of the Reactor n 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 sufficisaOsater inventory in the steam generator at the time of the trip l to provide

in of at least 10 minutes before auxiliary feedwater is L

required to- vent degraded core cooling.

1 Local Power Density - High 1

The Local Power Density - High trip is provided to prevent the linear '

heat rate (kW/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any design bases anticipated operational occur-rence. The local power density is calculated in the reactor protective system utilizing the following information:

PALO VERDE - UNIT 3 B 2-4 j

_ _ _ _ _ -- _ _---_.-__.- _ ____ _ _ _ - _ l

L. ,

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 l incorporates uncertainties and dynamic compensation routines. These uncer- I tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit. CPC uncertainties l related to peak LPD are the same types used for DNBR calculation. Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.

DNBR - Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences. The DNBR - Low trip incorporates a low pressurizer pressure floor of 1860 psia. At this pressure l a DNBR - Low trip will automatically occur. The DNBR is calculated in the CPC utilizing the following information:

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

b. Reactor Coolant System pressure from pressurizer pressure measurement;

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

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

e. Reaster coolant mass flow rate from reactor coolant pump speed;

f. Core inlet temperature from reactor coolant cold leg temperature measurements.

PALO VERDE - UNIT 3 8 2-5 AMENDMENT N0. 18 k .__x._..____u_____m__ _ _ - _ _ _ - _ - _ -

.y JSAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES DNBR - Low (Continued)

The DNBR, the trip variable, calculated by the CPC incorporates various uncer-tainties and dynamic compensation routines to assure a trip is_ initiated prior-to violation of fuel design limits. These uncertainties and dynamic compensa-tion routines ensure that a reactor trip occurs when the calculated core DNBR is sufficiently greater than 1.24 such that the decrease in calculated core l

DNBR after the trip will not result in a violation of the DNBR Safety Limit. I CPC uncertainties.related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays. !

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

a. RCS Cold Leg Temperature-Low > 470*F

b. RCS Cold Leg Temperature-High 7 610*F 1

~

c. Axial Shape Index-Positive Not more positive than + 0.5 i

d. Axial Shape Index-Negative Not more negative than - 0.5 ;

e. Pressurizer Pressure-Low > 1860 psia

f. j Pressurizer Pressure-High -7 2388 psia

g. Integrated Radial Peaking ,

Factor-Low > 1.28 1

h. Integrated Radial Peaking i Factor-High < 7.00 -

i. Quality Margin-Low FO l Steam Generator Level - High The Steasdienerator Level - High trip is provided to protect the turbine from excessivermeisture carry over. Since the turbine is automatically tripped when Se 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 i steam isolation setpoint. Its functional capability at the specified trip l

setting enhances the overall reliability of the reactor protection system. j l

l l

PALO VERDE - UNIT 3 8 2-6 AMENDMENT NO. 18 )

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FIGURE 3.1-1A SHUT 00WN MARGIN VERSUS COLD LEG TEMPERATURE PALO VERDE - UNIT 3 3/4 1-2a AMENDMENT NO. 18 j

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-__-________________._.__.____m_ _ _ _ _ _ _ _ _ _ _ _

a MINIMUM TEMPERATURE FOR CRITICALITY LIMITIhG CONDITION FOR OPERATION 3.1.1.4 shall be greater The Reactor Coolar,t than or Systen equal to 552*F. lowest operating loop temperature (Teold)

APPLICABILITY: *) DES 1 and 2#*.

ACTION:

)

With a Reactor Coolant System operating loop temperature (Tcold) less than 552*F, restore T eold to within its Itait within 15 minutes or be in HOT ,

, STAND 8Y within the next 15 minutes.

l SURVEILLANCE REQUIREMENTS 4.1.1. 4 The Reactor Coolant System temperature (Teold) 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 eold is less than 557'F.

With K,ff greater than or equal to 1.0.

See Special Test Exception 3.10.5.

)I l

PALO VERDE - UNIT 3 3/4 1-6 MENOMENT N0. 9

l REACTIVITY CONTROL SYSTEMS i

l 3/4.1.3 MOVA8LE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION i

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 18 and 2*,

ACTION:

a. With one or mor 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 SHUT 00WN 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 at least HOT STANDBY 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 J 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 I CEA(s) is either: l

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

2. Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.2 is satisfied. After declaring the CEA(s) inoperable, operation in MODES 1 and 2 may continue pursuant to ,

the requirements of Specifications 3.1.3.6 and 3.1.3.7 provided: 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 and the THERMAL POWER level restrictions of Specifications 3.1.3.6 and 3.1.3.7 !

during subsequent operation.

See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE - UNIT 3 3/4 1-21 AMENDMENT NO.18 j l

w,__,_.____-_-_m.. --w------- ' - - - - - ' - - - -

r, 1

LIMITING CONDITION F'OR OPERATION'(Continued)

, ' ACTION: (Continued)-

b) The SHUTDOWN MARGIN requirement of Specification 3.1.1.2 4 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . 1 x

Otherwise, be in at least HOT STAND 8Y 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- z ment requirements, operation in MODES 1 and 2 may continue pursuant 1

'to the requirements of Specification 3.1.3.6.

q

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.

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

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

1 l

l l

3 PALO VERDE - UNIT 3 3/4 1-22 AMENDMENT NO.18

1 l

THIS PAGE INTENTIONALLY DELETED I

'l l

PALO VERDE - UNIT 3 3/4 1-23 AMENDMENT NO.18

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i 6 0 10 20 30 40 50 60 E TIME AFTER DEVIATION, MINUTES a

1 I

'WHEN CORE POWER IS REDUCED TO 55% OF RATED THERMAL POWER PER THIS LIMIT CURVE, FURTHER REDUCTION IS NOT REQUIRED l

l l r

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FIGURE 3.1-2A CORE POWER LIMIT AFTER CEA DEVIATION

I PALO VERDE - UNIT 3 3/4 1-24 AMENDMENT NO. 18 i

t 1

L____-________--

POSITION INDICATOR CHANNELS - OPERATING 1

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, 3.1.3.5, 3.1.3.6 and 3.1.3.7. Operation- l 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 Specifica-tion 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 Eachief 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 i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

I s

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

1 PALO VERDE - UNIT 3 3/4 1-25 AMENDMENT NO.18

l 1

. REACTIVITY CONTROL SYSTEMS POSITION INDICATOR. CHANNELS - SHUTDOWN l

' LIMITING CONDITION FOR OPERATION j l

l' .- 3.1. 3. 3 At least one CEA Reed Switch Position Transmitter indicator channel shall be 0PERABLE for each shutdown, regulating or part-length CEA not fully inserted.

APPLICABILITY: MODES 3*, 4*, and 5*. .

l ACTION: ' '

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

.i l

SURVEILLANCE REQUIREMENTS

-l t

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

1 l

l I

l l

With the reactor trip breakers in the closed position.

PALO VERDE - UNIT 3 3/4 1-26

CEA DROP TIME LIMITING CONDITION FOR OPERATION o

.{

3.1.3.4 The individual full-length (shutdown and regulating) CEA drop time, from a fully withdrawn position, shall be less than or equal to 4 seconds from when the electrical power is interrupted to the CEA drive mechanism until the !

CEA reaches its 90% insertion position with: 1

a. Tcold greater than or equal to 552'F, and

b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2. !

ACTION:

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

1 SURVEILLANCE REQUIREMENTS ,

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

f

a. For all CEAs following each removal and reinstallation of the reac- I tor vessel head' !

b. For specifically affected individual CEAs following any maintenance !

on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and

c. At least once per 18 months. !

I l

l l l l l l PALO VERDE - UNIT 3 3/4 1-27 l

- - - - _ _ _ - - - _ _ _ _ - - _ - - - _ _ _ _ 2

j

"' ' A

SHUTDOWN CEA INSERTION LIMIT.

i LIMITING CONDITION FOR OPERATION

'3.1.3.5 'All shutdown CEAs'shall be withdrawn to at least 144.75 inches.

APPLICA8IllfY: MODES-1 and 2*#. I q

-ACTION:

With a maximum of one shutdown CEA withdrawn to less-than 144.75' inches',-

except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour'either:

a. Withdraw the CEA to at 1 east 144.75 inches, or

b. Declare the CEA inoperable and comply with. Specification 3.1.'3.1.

l SURVEILLANCE REQUIREMENTS l

'4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least' 144.75 inches:

i

a. Within 15 minutes prior to withdrawal of any CEAs in regulating !

groups during an approach to reactor criticality, and !

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter except during time intervals 1' when both CEAC's'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 .

J

See Special Test Exception 3.10.2. !

With K,ff greater than or equal to 1.

PALO VERDE - UNIT 3 3/4 1-28 AMENDMENT NO.18

_ _ = -_ _ _ _ -- - _ - . . _ - _ _

1 REGULATING CEA INSERTION LIMITS <

LIMITING CONDITION FOR OPERATION 3.1.3.6 The regulating CEA groups shall be maintained within the following limits:

a. One or more CEAC's operable:

1. The regulating CEA groups shall be limited to the withdrawal l sequence, and to the insertion limits ## shown on Figure 3.1-3 1 when the COLSS is'in service or shown on Figure 3.1-4 when the COLSS is not in service. The CEA insertion between the Long i Term Steady State Insertion Limits and the Transient Insertion Limits is restricted to:

a) Less than or equal to 5 Effective Full Power Days per 30 I Effective Full Power Day interval, and b) Less than or equal to 14 Effective Fu11' Power Days per 18 !

Effective Full Power Months.

2. CEA insertion between the Short Term Steady State Insertion Limits and the Transient Insertion Limits shall be restricted to $,4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval.

b. Both CEAC's Inoperable (with or without COLSS in service):

Regulating CEA group 5 may be inserted no further than 127.5 inches withdrawn which is the Transient Insertion Limit when both CEAC's are. inoperable.

I t

Regulating CEA groups which are excluded by these insertion limits must be maintained fully withdrawn > 144.75 inches, which is the Transient Insertion Limit, except for surveillance testing pursuant to Specifica-tion 4.1.3.1.2.

APPLICABILITY: MODES 1* and 2*#. .

l ACTION:

a. Witle t,he regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the regulating CEA groups to within the limits, or

2. Reduce THERMAL POWER as follows:

a) One or more CEAC's Operable

1) Reduce THERMAL POWER to less than or equal to that ,

fraction of RATED THERMAL POWER which is allowed by '

the CEA group position using Figures 3.1-3 or 3.1-4, 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 .

PALO VERDE - UNIT 3 3/4 1-29 AMENDMENT NO.18 j

i

-REGULATING'CEA INSERTION LIMITS

~ LIMITING CONDIT10N'FOR OPERATION'(Continued)

ACTION: - (Continued).

b) Both CEAC's' Inoperable l

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

b. ~

With the regulating.CEA' groups inserted between the Long. Term Steady.

~ State Insertion Limits and the Transient Insertion Limits.for inter-vals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per 18. Effective Full Power Months, either:

.- l . Restore the regulating groups to within the Long Term' Steady State Insertion Limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

2. Be in'at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f

.c. With the regulating CEA groups inserted between the Short Term Steady State Insertion Limits and the Transient Insertion Limits-for intervals > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval, operation may proceed-provided any subsequent increase in THERMAL POWER is restricted to

< 5% of RATED THERMAL POWER per hour.

SURVEILLANCE REQUIREMENTS-4.1.-3.6 The position of each regulating CEA group shall be determined to be within the Transient' Insertion Limits at least once per 12' hours except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, or both CEAC's are inoperable,.then verify'the CEA group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . 'The accumulated times during which the regulating CEA groups are inserted beyond the Long Ters Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i

See Special Test Exceptions 3.10.2 and 3.10.4.

.{

With K greater than or equal to 1.

1. A reac h power cutback will cause either (Case 1) Regulating Group 6 or I Regulating Group 4 and 5 to be dropped with no sequential insertion of ,

additional Regulating Groups (Groups 1, 2, 3, and 4) or (Case 2) Regulating '

Group 5 or Regulating Groups 4 and 5 to be dropped with all or part of the L remaining Regulating Groups (Groups 1, 2, 3, and 4) being sequentially inserted. In either case, the Transient Insertion Limit and the withdrawal .I sequence of Figure 3.1-3 or Figure 3.1-4 can be exceeded for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . l I'

PALO VERDE - UNIT 3 3/4 1-30 AMENDMENT NO. 18 !

]

4, . .

j 1

[. REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS '

L

' LIMITING CONDITION FOR OPERATION-3.1. 3. 7 The part length CEA groups shcIl be maintained within the following~

limits-with COLSS in service or out of service:

a. One-or more CEACs OPERABLE

- 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:

-1. .s 7 EFPD per 30 EFP0 interval, and

2. 1 14 EFPD per calendar year.

b. Both' CEACs INOPERABLE The part length CEA groups 'must be maintained fully withdrawn (~> 144.75 inches) which.is the Transient Insertion Limit when'both CEAC's are inoperable.

APPLICABILITY: MODES 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 groups to within the limits,' or

2. Reduce THERMAL POWER as follows:

a) One or more CEACs OPERABLE

1) Reduce THERMAL POWER to less than or equal to that l i

fraction of RATED THERMAL POWER which is' allowed by {

the PLCEA group position using Figure 3.1-5, or

2) Be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ;

b) Both CEACs INOPERABLE Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

PALO VERDE - UNIT 3 3/4 1-30A AMEN 0 MENT NO.18 A-_ a__._ -__-2m__. _._____..!__________-__

s f REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued)

b. With the part length CEA groups inserted between the Long Term Steady State Insertion Limit and the Transient Insertion Limit for j intervals > 7 EFPD per 30 EFPD interval or > 14 EFPD per calendar l year, either:

j j

1. Restore the part length groups within the Long Term Steady {

State Insertion Limit within two hours, or

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

SURVEILLANCE REQUIREMENTS I

4.1.3.7 The positions of the part length CEA groups 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 except during time intervals when both CEACs are inoperable, then verify the part length CEA group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

t 1 .

l 1

i l

l l

See Special Test Exceptions 3.10.2 and 3.10.4.

PALO VERDE - UNIT 3 3/4 1-30B AMENDMENT N0.18

1

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SHORT TERW STEADY I fl i STATO NSERTION LtWIT i l

O r r r r --P T t -7 7 : 9 1 l

',r GROUP 5 e 10F LONG TERM 'I STEADY.m - l STATE NSERTm815#T

' I I I I I I t iI iT t i t 1iI i i i I

~":e 8 8-8 R 8

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d' d d 6 $*6R.

o.6 6 o 8 6 FRACTION OF RATED THERMAL POWER PALO VERDE - UNIT 3 3/4 1-31 AMENDMENT NO. 18 i

l

)

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I i l l l l I I i i '

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8 TRANSIENT INSERTION LIMIT % -

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b ~ 83 8 I

$ C gg i l I l i i l ll i 1' 1 1 i l i i l i i i l\ /l 1 -l _g _g y 6 a t "

l / I dv l

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g i

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, / GROUP 5 9 60" 3 -8 *

/ SHORT TERM STEADY 1

i f STATE ,1NSERTIO,N LIMIT ,

c i S

r r , 1 r r c r ri y I j GROUP 5 0 108" LONG TERM STEADY g;g -

E u

I i ST, ATE INSERTION LIMIT

, i i e i i e i e i i l I I

_g l a

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a 6

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s d l FRACTION OF RATED THERuAL POWER

{

PALO VERDE - UNIT 3 3/4 1-32 AMENDMENT NO. 18

t i ,- ,;

I g g g i g 0 g

" l

- g l

- 20

- 30 '

~

- 40 g

=

- 50 ,5

.I

" w

- 60 W -

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I TRAN51Dif NSERTION LMT (75.0 WCHts 3 so l 5

so UNACCEPTABLE RESTRICTED I CPERAT10N -

OPcRAT10N E

- 20 $

, k *

- no =

=

LONO TERM $7EADY STATE NSERTION LAAT -

20 caz.s NCHesi

- no

~

= ido f I f f f I I ( l 20 g 8 3 2 S S S R

q 8 a 6 6 6 6 6 6 6 o o a FRACTION OF RATED THERWAL power J

FIGURE 3.1-5 4

I PART LENGTH CEA INSERTION LIMIT VS THERMAL POWER j i

PALO VERDE - UNIT 3 3/4 1-33 AMENDMENT NO. 18 ]

I

- 1

- i

_ - - - - - - - _ - - - - - - - - - - - - - - - 1

. l 1

3/4.2 POWER DISTRIBUTION LIMITS 1

3/4 2.1 LINEAR HEAT RATE '

I 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:

l

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

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

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

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

b. Reduce THERMAL POWER to less than or equal to 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 previsions of Specification 4.0.4 are not applicable.

4.2.1.2 The Ifnear heat rate shall be determined to be within its limit when l 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 th3 linear heat rate, as indicated on any OPERABLE Local Power Density channel, is within its limit.

4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating limit based on linear heat rate. l PALO VERDE - UNIT 3 3/4 2-1 AMEN 0 MENT NO. 18

m e

1 s

, _,: u i

t POWER DISTRIBUTION LIMITS 3/4.2.2 PLANAR RADIAL PEAV.ING FACTORS - F

. LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL. PEAKING FAC ORS (F" ) shall be less.than~or equal to the PLANAR RADIAL PEAKING FACTORS (F in the Core Operating Limit Supervisory System (COLSS) and inProtection the C fe)Calculators use5Y (CPC).

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

ACTION.

g e

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'

~

appliedtoplanargadiglpeakingbyafactorequivalenttogreater than or_ equal to F /F margintotheCOLS5Y opIfa.andrestrictsubsequentopegatignsothata.

ting limits of at least [(Fxy/F*Y).- 1.0)-

x 100% is maintained; or c

b. Adjust the;affected PLANAR RADIAL PEAKING FACTORS (F used in the COLSS and'CPC to a value greater,than or equal m to thiY)easure PLANAR RADIAL PEAKING FACTORS (Fyy) or

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

POWER.

! SURVEILLANCE REQUIREMENTS l.

\

4.2.2.1 The provisions of Specification 4.0.4 are-not applicable.

4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F" obtained by_using theincoredetectionsystem,shallbgdeterminedtobe*Te)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 I

b. At least once per 31 Effective Full Power Days.

See Special Test Exception 3.10.2.

)

I J

i PALO VERDE - UNIT 3 3/4 2-2 l

{' _____-___-________A

y.

c 1.0 4

0.9 A

- f- O.8 M

U T

H 0.7 R MWQN CP L

.P 0.6 0

W <

E R 0.5 7

.I

~L 0.4 i l

T l

l F

[0.3 C ,

l T l-f0.2 N

0.1 ns30N CP 0.0 .

20 -

39 -

40 50 60 70 80 90 100 ,)

PERCENT OF RATED THERMAL POWER FIGURE 3.2-1A AZIMUTHAL POWER TILT LIMIT VS THERMAL P0kIR (COLSS IN SERVICE)

PALO VERDE - UNIT 3 3/4 2-4a AMENDMENT NO. 18 l

g + y L#

g 4 S . POWER' DISTRIBUTION. LIMITS 3/4.2.~4 DN8R MARGIN !

' LIMITING. CONDITION FOR OPERATION 3.2.4-',The DNBR margin'sh'll a be maintained by-one of the following methods: I

a. . Maintaining COLSS calculated core power ~1ess.than or equal to.COLSS' o,

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

b. Maintaining COLSS calculated core power.less than or equa'l to COLSS' calculated core power operating limit based on DNBR decreased by the i 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 s

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

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

ACTION: ;

With the DNBR not being maintained:

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 increase the DNBR'to within the limits and either:

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

b. Reduce THERMAL POWER to less than or equal to 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 MAUIREMENTS 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 ON8R channel, is within the limit shown on Figure 3.2-2 or Figure 3.2-2A.

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

PALO VERDE - UNIT 3 3/4 2-5 AMENDMENT NO.18

e I 4 4 +

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in

_N m

-s e s

E E

w i!!! 1'

-R E l

1 .

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m a COLSS ONER POWER OPERATING LIMIT REDUCTIOW

(% OF RATED THERMAL POWER)

FIGURE 3.2-1 COLSS DNBR POWER OPERATING LIMIT ALLOWANCE FOR BOTH CEACs INOPERABLE PALO VERDE - UNIT 3 3/4 2-6 AMENDMENT NO. 18 1

COLS5 OUT OF RRVICE ONBR LIMIT LINE 2*1

1. 1 4

20 - ACCEPTABLE

~

OPERATION MINIMUM 1 CEAC OPERABLE 1.9 -

(1.1.87) (.2,I.87) kz 1.8 -

E .

M w ( .2.1.75)

UNACCEPTABLE OPERATION 1.6 -

t i f i 1

-0.3 -0.2 ' -0.1 2.9 0.1 0.2 2.3 CORE AVERAGE ASI FIGURE 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs OPERABLE)

PALO VERDE - UNIT 3 3/4 2-7 AMEN 0 MENT NO. 18 m

COLSS OUT OF SERVICE ONBR LIMIT LINE' 1

2.4 1

l ACCEPTABL'E OPERATION

. 2.3 -

CEACs INOPERABLE

(.05.2.30) L2,2.301 -

g -

-2.2 -

~

2

/ -

I 2.1 -, ( .2,2.13)

( ). -

Si UNACCEPTABLE OPERATION 2.0 -

i U

1.9 -- ' --

l l i ,

-0.3 -0.2 -fJ,1 0.0 0.1 0.2 0.3 CORE AVERAGE ASI FIGURE 3.2-2A DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE, CEACs INOPERABLE)

PALO VERDE - UNIT 3 3/4 2-7a AMENDMENT NO.18 l l

i

.- 1

=

)

')

POWER DISTRIBUTION LIMITS i

3/4.2.5 RCS FLOW RATE LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reactor Coolant System total flow rate shall be greater than or equal to 155.8 x 108 lbm/hr.

l APPLICA8ILITY: 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 . ;

i 4

i l

PALO VERDE.- UNIT 3 3/4 2-8 AMEN 0 MENT NO.18 l

q

. ; : e 3. :

N '

)

. POWER DISTRIBUTION' LIMITS

' 3/4.2.7 AXIAL SHAPE INDEX l t:

LIMITING CONDITION FOR OPERATION 3.2.7 The core. average AXIAL SHAPE INDEX (ASI) shall be. maintained within the

<- following limits:

a. COLSS OPERABLE-

' 0.28 5 ASI 11 0 28

'. b. COLSS'0UT OF SERVICE.(CPC) i

.-0.20 $ ASI 5 + 0.20 APPLICABILITY: MODE'l 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 'i I

4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within its~

limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months using the COLSS or any OPERABLE Core Protection Calculator channel.

"See Special Test Exception 3.10.2.

i

'PALO VERDE - UNIT 3 3/4 2-11 l

_ _ - -- - 4

POWER DISTRIBUTION LIMITS 3/4.2.8 PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION l

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

APPLICABILITY: MODES 1 and 2*.

' ACTION:

With the pressurizer pressure outside its above limits, restore the pressure

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

9 i

SURVEILLANCE REQUIREMENTS '

l 4.2.8 The pressurizer pressure shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i

(

See Special Test Exception 3.10.5 a i

?

PALO VERDE - UNIT 3 3/4 2-12 AMENDMENT NO. 18 L_________-_____-___ !

e TABLE 3.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION a ACTION STATEMENTS l

l

3. Steam' Generator Pressure - Steam Generator Pressure - Low-Low Steam Generator Level 1-Low (ESF)

Steam Generator Level 2-Low (ESF)

4. Steam Generator Level - 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)

DNBR - Low (RPS)

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

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

ACTION 4 -

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

ACTION 5 -

With the number of channels OPERABLE one less than required by'the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the reactor' trip breaker of the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, the trip breaker associated with the inoperable channel may be closed for up to I hour for surveillance testing per Specification 4.3.1.1.

ACTION 6 -

a. With one CEAC inoperable, operation may continue for up to 7 days provided that the requirements of Specification 4.l.3.1.1 are met. After 7 days, operation may continue provided that ,

the conditions of Action Item 6.b are met. i

b. With both CEACs inoperable operation may continue provided that:

i I

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

J TABLE 3.3-1 (Continued)

REACTOR PROTECTIVE INSTRUMENTATION. I ACTION STATEMENTS 1.. -Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the DNBR margin required by Specifica-tion 3.2.4b (COLSS in service) or 3.2.4d (COLSS out of service) is satisfied and the Reactor Power Cutback j System is disabled, and l 1

2. -Within 4' hours:

i a) All full-length and part-length CEA groups must be withdrawn within the limits of Specifica-tions 3.1.3.5, 3;1.3.6b,-and'3.1.3.7b, except- i !

during surveillance testing pursuant to the requirements of. Specification 4.1.3.1.2. - Specifi-cation 3.1.3.6b allows CEA Group'5 insertion to no further than 127.5 inches withdrawn.

b) The "RSPT/CEAC Inoperable" addressable constant.

in-the CPCs is set to indicate that both CEAC's are. inoperable.

c) The Control Element Drive Mechanism Control 1 System (CEDMCS) is maintained in the "placed in and subsequently Standby" mode except during l CEA motion permitted by Specifications 3.1.3.5, 3.1.3.6b and 3.1.3.7b, when the CEDMCS may be operated in either the " Manual Group" or " Manual Individual" mode.

t

3. CEA position surveillance must meet the requirements i of Specifications 4.1.3.1.1, 4.1.3.5, 4.1.3.6 and ;

4.1.3.7 except during surveillance testing pursuant to Specification 4.1.3.1.2.

ACTION ~7 -

With three or more auto restarts, excluding periedic auto restarts (Code 30 and Code 33), of one non bypassed calculator {

during a.12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL Fur 4CTIONAL 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 OPERABLE chanr.els one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open an affected reactor trip breaker within the next hour, f

i PALO VERDE - UNIT 3 3/4 3-8 AMENDMENT NO. 18 ,

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, : 3/4'.10 SPECIAL TEST EXCEPTIONS

3/4.10.1~ SHUTDOWN MARGIN AND K N CEA WRN TESTS l LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN and K' requirements of Specification 3.1.1.2 may N-1 l be~ suspended for measurement of CEA worth and i,hutdown 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, immedi-ately initiate and continue boration at greater than or equal to 26 gpa of a solution containing greater than or equal to 4000 ppm boron by or its equivalent Specification untilare 3.1.1.2 therestored.

SHUTDOWN MARGIN and yK 1 required

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

l SURVEILLANCE REQUIREMENTS 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 . l l

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.

4.10.1.3 When in MODE 3 or M005 4, the reactor shall be determined to be 4 suberitical by at least the reactivity equivalent of the highest estimated CEA i worth or the reactivity equivalent of the highest estimated CEA worth is avail- i able for trip insertion from OPERABLE CEAs at least once per 2 nours by con-sideration of at least the following factors: l

a. Reacter Coolant System boron concentration, i

b. CEA position,

c. Reactor Coolant System average temperature,

d. Fuel burn 6p based on gross thermal energy generation,

e. Xenon concentration, and I

f. Samarium concentration.

I Operation in MODE 3 and MODE 4 shall be limited to 6 cor.secutive hours.

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

PALO VERDE - UNIT 3 3/4 10-1 AMENDMENT NO. 2 L--__-____________

{ ,

l SPECIALTETTEXCEPTIONS 1

3/4.10.2- MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 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, q 1

3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, and the Minimum Channels OPERABLE requirement '1 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 Ifmits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.

APPLICA8ILITY: MODES 1 and 2.

~

u ACTION:

With any of the limits of Specification 3.2.1 being exceeded while 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, and the. Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended, either:

l{ .

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 PHYSICE TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.L 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 plater,u.

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, 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 3 3/4 10-2 AMENDMENT NO. 18

q. ,.

SPECIAL TEST EXCEPTIONS' l 1

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 i; Tables 2.2-1 and.3.3-1 may be suspended during the performance of-startup PHYSICS TESTS, provided: J

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and -]

b. The reactor trip setpoints of th'e 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.

APPLICABILITY: During startup PHYSICS TESTS.

ACTION:

i With the THERMAL POWER greater than 5% of RATED THERMAL POWER or with less than j 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 ar equal to 5%

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

4.10.3.2 Each logarithmic and vtriable 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. hours.

l l

PALD VERDE - UNII 3 3/4 10-J ,

a

f

I

.SPECIAL TEST EXCEPTIONS )

e

'3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT-COLD LEG TEMPERATURE LIMITING CONDITION FOR OPERATION l

3.10.4 The requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7 and 3.2.6 may be suspended during the performance of PHYSICS TESTS to determine ll l i

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.

t A. CTION:

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With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7 and 3.2.6 are. suspended, either: l4

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 L

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, 3.1.3.5, 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. ,

l 4.10.4.2 The linear heat rate shall be determined to be within the limits of f Specification 3.2.1 by monitoring it continuously with the Incore Detector 1 Monitcring System pursuant to the requirements of Specification 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the requirements )i of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7 and/or 3.2.6 are suspended.

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PALO VERDE - UNIT 3 3/4 10-4 AMENDMENT NO. 18

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I, REACTIVITY CONTROL SYSTEMS I '

BASES MOVABLE CONTROL ASSEMBLIES (Continued) and LS$$ 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 l

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 inchas inserted over the fuel cycle. This cycling will distribute the possible guide tube waar over a larger area, thus minimizing any effects. To acccamodate this l' programmed insertion schedule, the fully withdrawn position was redefined, in some cases, to be 144.75 inches or greater.

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

, 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 pesks 4 due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions i

I PALO VERDE - UNIT 3 8 3/4 1-5

'REACTIVfTY CONTROL SYSTEMS l

' BASES !

i MOVABLE CONTROL ASSEMBLIES (Continued) i and. load maneuvering.' Analyses are performed based on the expec.ted' mode of ,

operation:of the NSSS (base load maneuvering, etc.) and from these analyses i CEA insertions are determined.and a consistent set of radial peaking factors defined. The Long Term Steady State and Short Term Insertion Limits are deter-i]

mined based upon'the. assumed mode of operation used'in the analyses and provide' a means 'of preserving the assumptions on CEA insertions used. The limits speci-fied serve'to limit the behavior of the radial peaking factors within the bounds' 'q 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 1 specified for the plant which'has been designed for primarily base loaded 1 operation but which has the ability to accommodate a limited amount of load I maneuvering. l i.

The Transient Insertion Limits of Specifications 3.1.3.6 and 3.1.3.7 and !

the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that (1) the I

. minimum SHUTDOWN MARGIN is maintained, and-(2) the potential effects of.a CEA '

ejection accident are limited to acceptable levels. Long-term operation at the Transient Insertion Limits is not permitted since such operation could have:

. effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial peaking factors.

The PVNGS CPC and COLSS systems are responsible for the safety and monitoring functions, respectively, of the reactor core. COLSS monitors the DNB-Power Operating Limit (POL) and various operating 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- I acceptable fuel damage.

1 1

The COLSS reserves the Required Overpower Margin (ROPM) to account for.

Loss of Flow (LOF) and CEA misoperation transients. When the COLSS is Out of Service-(C005), the monitoring function is performed via the CPC. calculation of 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.

The reduction of the CEA deviation penalties in accordance with the CEAC (Control Element Assembly Calculator) sensitivity reduction program has been pr.rformed. This task involved setting many of the inward single CEA deviation pe alty factors to 1.0. An inward CEA deviation event in effect would not be i accompanied by the application of the CEA deviation penalty in either the CPC DNB and LHR (Linear Heat Rate) calculations for those CEAs with the seduced penalty factors. The protection for an inward CEA deviation event is thus accounted for separately. 1 1

PALO VERDE - UNIT 3 8 3/4 1-6 AMENDMENT NO. 18 L . .

REACTIVITY. CONTROL SYSTEMS BASES MOVABLE' CONTROL ASSEMBLIES (Continued) i 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 tyt;th CEA maneuvering is restricted in l !

accordance 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 3 8 3/4 1-7 AMENDMENT NO. 18 l

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l' 3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200 F.

Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution und 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 l limit corresponding to the allowable peak linear heat rate. Reactor operation '

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

The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to the operator. A COLSS alarm is annunciated in the event that the ,

core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady-state opera-tion. Normal reactor power transients or equipment failures which do not require a reactor trip may result in this core power operating limit being exceeded. In the event this occurs, COLSS alarms will be annunciated. If the event which causes the COLSS limit to be exceeded results in conditions which !

approach the core safety limits, a reactor trip will be initiated by the Reactor l Protective Instrumentation. The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability / i 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. l Parameters required to maintain the operating limit power level based on linear heat rate, margin to DNB, and total core power are also monitored by the CPCs. Therefore, in the event that the COLSS is not being used, operation within .

the linear heat rate limit can be maintained by utilizing any operable CPC channel. The above listed uncertainty and penalty factors plus those associated with the CPC startup test acceptance criteria are also included in the CPCs.

I 18 PALO VERDE - UNIT 3 8 3/4 2-1 AMEN 0 MENT NO.

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P_0WER DISTRIBUTION LIMITS BASES-3/4.2.2 PLANAR RADIAL PEAKING FACTORS- -

Limiting the' values of the PLANAR RADIAL PEAKING FACTORS-(F c) used inEthe JCOLSS and CPCs to values equal to or greater than the measured PLANAR RADIAL PEAKING. FACTORS (F,y") provides: assurance that the limits calculated by C01,55 and the CPCs remain valid.. Data from the incore detectors are used for-determining the measured PLANAR RADIAL PEAKING FACTORS. A minimum ccre power at 20% of RATED 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 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 1 exceeding 70% of RATED THERMAL POWER provides additional ~ assurance that the core was properly loaded.- j.

i 3/4.2.3 AZIMUTHAL POWER TILT - T, The limitations on the AZIMUTHAL POWER TILT are provided to ensure that

' design safety margins are maintained. An AZIMUTHAL POWER TILT greater than the limit in Figure 3.2-1A with COLSS in service or 0.10 with COLSS out of service is not' expected and if it should occur,' operation is restricted to only l those conditions required to identify the cause of the tilt. The tilt is normally calculated by COLSS. A minimum core power of 20% of RATED THERMAL POWER is assened by the CPCs in its input to COLSS for calculation of AZIMUTHAL POWER TILT. The 205 RATED THERMAL POWER threshold is due to the neutron flux detector system being inaccurate below 20% core power. Core-noise ieval at low power is too'large to obtain usable detector readings. The surveillance requirements specified when COLSS.is out of service provide an

, acceptable sec4 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 Al @ L POWER TILT is equal to (Ptilt/Puntilt)-1.0 where:

AZIMITML 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:

tilt/Puntilt = 1 + Tqg cos (0 - So)

P where:

Tq is the peak fractional tilt amplitude at the core periphery g is the radial normalizing factor }

8 is the atinuthal core location So is the azimuthal core location of maximum tilt PALO VERDE - UNIT 3 8 3/4 2-2 AMEN 9 MENT NO. M

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.' POWER DISTRIBUTION LIMITS BASES

" AZIMUTHAL POWER TILT - Tq (Continued)-

1 Ptig/Puntilt is the ratio of the power at a core location in the presence i of a tilt to the power at that location with no tilt.

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

3/4.2.4 DNBR MARGIN The limitation on~DNBR as a function of AXIAL SHAPE INDEX represents a conservative envelope of operating conditions consistent with the safety analy-sis assumptions which have been analytically demonstrated adequate..to main-tain an acceptable minimum DNBR throughout all anticipated operational occur-rences. Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a ]?

loss of flow transient.

Either of theLtwo 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-l: limits. The COLSS performs this function by continuously monitoring the core

[ power distribution and calculating a core operating lim _it 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' 1

' (based on the minimum DNBR Limit) are conservative with respect to,the' actual l core' power limit. These penalty factors are determined from the. uncertainties associated with planar radial peaking measurement, engineering heat flux,. state parameter measurement, software algorithm modelling, computer processing, rod bow, and core power measurement.

Parameters required.to maintain the margin to DN8 and total core power are also monitored by the CPCs. Therefore, in the event that the COLSS is not-being used, operation within the limits of Figures 3.2-2 and 3.2.2a'can be l maintained by~ utilizing a predetermined DNBR as a function of AXIAL SHAPE INDEX ;

j and by monitoring the CPC trip channels. The above listed uncertainty and l 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 lausi at low power is too large'to obtain usable detector readings.

A DN8R penalty factor has been included in the COLSS and CPC DNBR calcula- 3 tions to accommodate the effects of rod bow. The amount of rod bow in each i assembly is dependent upon the average burnup experienced by that assembly. l Fuel assemblies that incur higher average burnup will experience a greater magnitude of rod bow. Conversely, lower burnup assemblies will experience less rod bow. In design calculations, the penalty for each batch required to com- l pensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC is then determined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.

PALO VERDE - UNIT 3 8 3/4 2-3 AMENDMENT NO.18 L

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l POWER DISTRIBUTION LIMITS-l 5

BASES j L 3/4.2.5 RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow j rate is maintained, at or above the minimum value used in the safety analyses. 1 The minimum value used in the safety analysis is 95% of the design flow

rate (164.0 x 108 lbm/hr) or 155.8 x 108 lbm/hr. The actual-RCS flow rate is !

determined 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. j 3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor '

coolant cold leg temperature is maintained within the range of values used in the safety analyses.

3/4.2.7 AXIAL SHAPE INDEX 'i i

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.

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l PALO VERDE - UNIT 3 8 3/4 2-4 AMENDMENT NO.18

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3/4.3 INSTRUMENTATION BASES

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f- 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES l L ACTUATION SYSTEM INSTRUMENTATION f 1 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 i

when the parameter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient _

redundancy is maintained to permit a channel to be out of service for testing or. maintenance, and (4) sufficient system functional capability is available from diverse parameters.

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

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

The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications of power level, RCS flow rate, axial flux shape, radial peaking factors and CEA deviation-penalties. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the +

CPCs is unlikely.

The design of the Control Element Assembly Calculators (CEAC) provides reactor protection in the event one or both CEACs become inoperable. If one CEAC is in test or inoperable', verification of CEA position is performed at least every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the second CEAC fails, the CPCs in conjunction with plant Technical Specifications will use DNBR and LPD penalty factors and increased DNBR and LPD margin to restrict reactor operation to a power level that will ensure safe operation of the plant. If the margins are not maintained, a reactor trip will occur.

The value of the DNBR in Specification 2.1 is conservatively compensated ;

for measurement uncertain. ties. Therefore, the actual RCS total flow rate ,

determined by the r e tor coalant pump differential pressure instrumentation or i by calorimetric calculations does not have to be conservatively compensated for measurement uncertainties.

PALO VERDE - UNIT 3 8 3/4 3-1 AMENDMENT NO. 18

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INSTRUMENTATION d

BASES a 1

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

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each '

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

No credit was taken in the analyses for those channels with response times indicated as not applicable. The response times in Table 3.3-2 are made up of the time to generate the trip signal at the detector (sensor response time) and i thetrip or time for the delay signal to interrupt power to the CEA drive mechanism (signal time). ,

l1 Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the 4

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.

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. i' 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient I 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 instrementa- j tion for the site is listed in Table 3.3-7.

l 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 ;

l doses to the public as a result of routine or accidental release of radioactive l

l PALO VERDE - UNIT 3 8 3/4 3-2 AMENDMENT NO. 18 j i

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INSTRUMENTATION ,

1 BASES- .

METEOROLOGICAL-INSTRUMENTATION'(Continued)

. 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.ue' measured by the meteorological instrumentation.

3/4.3.3.5 REMOTE SHUTDOWN SYSTEM The.0PERABILITY'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 consistently with General Design Criterion 19 of 10 CFR Part 50.

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 instruments-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.

3/4.3.3.6 POST-ACCIDENT MONITORING INSTRUMENTATION The OPERSILITY 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&B 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 PALO VERDE - UNIT 3 B 3/4 3-3 AMENDMENT NO.18

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INSTRUMENTATION BASES L !

POST-ACCIDENT MONITORING INSTRUMENTATION (Continued)

)

i inl Table 3.3-13. The containment hydrogen monitors are in Specifica-tion 3/4.6.4.1. The Post-Accident Sampling System (RCS coolant) is in Table 3.3-6.

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

and the Core Exit Thermocouple (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.

the plant to ColdThese are not required by the accident analysis, nor.to bring Shutdown. '

l 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 i

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 refueling outage as soon as reasonably possible. j

.j 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 system and avoid or mitigate damage to primary system components. The allowable out-of-service times and surveillance requirements are consistent with the i recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981. !

3/4.3.3.8 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radiesctive gaseous effluent instrumentation is provided to monitor and control, as aplicable, the releases of radioactive 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 parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation 1 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.

PALO VERDE - UNIT 3 B 3/4 3-4 AMENDMENT NO. 18

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INSTRUMENTATION BASES RADI0 ACTIVE GASEOUS' EFFLUENT MONITORING INSTRUMENTATION (Continued) i There are two separate radioactive gaseous effluent monitoring systems:

the low range effluent monitors for normal plant radioactive gaseous effluents ]

1 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 the 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 1,ow range monitors. t f

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PALO VERDE - UNIT 3 B 3/4 3-5 AMENDMENT NO. 18 L_-__-____-_-__-_--_______ _ _ _-_.

j 3/4.4 REACTOR COOLANT SYSTEM /

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION <

The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.24 l during all normal operations and anticipated transients. In MODES 1 and 2 '

with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat removal l capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires that two shutdown cooling loops be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.

The operation of one reactor coolant pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification, and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 4000 gpm will circulate one equivalent ,

l Reactor Coolant System volume of 12,097 cubic feet in approximately 23 minut: 4 The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a reactor coolant pump in MODES 4 and 5, with ,

one or more RCS cold legs less than or equal to 255*F during cooldown or 295 F during heatup are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures.

3/4.4.2 SAFETY VALVES The presswH zer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve a minimum of 460,000 lb per hour of satur: tea steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur dering shutdown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

i I PALO VERDE - UNIT 3 8 3/4 4-1 AMEN 0 MENT NO.18

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REACTOR COOLANT' SYSTEM BASES SAFETY VALVES (Continued)

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.

The combined relief capacity of'these-valves is sufficient to limit the system pressure to within its-Safety Limit of 2750 psia following a completeLloss of turbine generator load:while operating at RATED THERMAL POWER and assuming no

. reactor trip until the.first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e...there is no direct reactor trip on the loss of turbine) and also assuming no operation of the steam dump valves.

Demonstration of-the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI .

of.the ASME Boiler and Pressure Vessel-Code.

3/4.4.3 PRESSURIZER An OPERABLE pressurizer provides pressure control for the Reactor Coolant System during operations with both' forced reactor coolant flow and with natural circulation flow. The minimum water level in-the pressurizer assures the pressurizer heaters, which are required to achieve and maintain pressure control, remain covered with water to prevent failure, which could' occur if the heaters were energized uncovered. The maximum water level.in the pressurizer ensures ~

that this parameter.is maintained within the envelope of operation assumed in i-the' safety analysis. . The' maximum water level also ensures that the RCS is'not a hydraulically solid system and..that a steam bubble will be provided to i accommodate pressure ~ surges during operation. The steam bubble also protects the pressurizer code safety valves against water relief. The requirement to verify that on an Engineered Safety Featur'es Actuation test signal concurrent with a. loss-of-offsite power the pressurizer heaters are automatically shed .

from the emergency power sources is to ensure that the non-Class lE heaters do not reduce the reliability of or overload the emergency power source. .The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability to control Reactor Coolant System pressure and establish and H maintain natural circulation. !

The auxiliary pressurizer spray is required to depressurize the RCS by cce'.

ing the pressurizer steam space to permit the plant to enter shutdown cooling.

The auxiliary pressurizer spray is required during those periods when normal pressurizer spray is not available, such as during natural circulation and during i the later stages of a normal RCS cooldown. The auxiliary pressurizer spray also ,

distributes boron to the pressurizer when normal pressurizer spray is not avail-able. Use of the auxiliary pressurizer spray is required during the recovery from a steam generator tube rupture and a small loss of coolant. accident. j 4.

6 PALO VERDE - UNIT 3 B 3/4 4-2

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i f Iog UNITED STATES j y 3 .e ( p, NUCLEAR REGULATORY COMMISSION j

.", j WA$HINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION )

RELATED TO AMENDMENT NO.18 TO FACILITY CPERATING. LICENSE NO. NPF-74 1

ARIZONA PUBLIC SERVICE COMPANY, ET AL. 1 i

PALO VERDE NUCLEAR GENERATING STATION, UNIT.NO. 3

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DOCKET.NO. STN 50 530 l i

1.0 INTRODUCTION

By letter dated December 14, 1988 (Ref. 1), the Arizona Public Service Company I (APS) on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edison Company, El Paso Electric Company, 3 Public Service Company of New Mexico. Los Angeles Department of Water and ' '

Power, and_ Southern California Public Power Authority (licensees), requested several changes to the Technical Specifications (Appendix A to facility Operating License No. NPF-74) for the Palo Verde Nuclear Generating Station, Unit 3 (PVNGS3), relating to Cycle 2 operation for PVNGS3. In support of both the Technical Specification changes and Cycle 2 operation, the licensees i submitted a Reload Analysis Report by letter dated December 27, 1988 (Ref. 2).

By letters dated April 26, 1989 (Refs. 3 and 4), the licensees also provided l corrections to the Reload Analysis Report necessitated by the revised '

end-of-cycle 1 termination burnup of 397 Effective Full Power Days, and infor-mation requested by the staff concerning the applicability of Reload Analysis Report references. The staff's evaluation of the reload analysis is presented <

in Sections 2.0 through 2.6 below. The evaluation of the specific changes to the Technical Specifications is presented in Section 3.0 below.

The reload will include 104 new Batch D assenblies, while 69 Batch A and 35 Batch 8 assemblies will be removed. All Batch C assemblies will remain for Cycle 2. Cycle 2 will operate at the rated core power of 3,800 MWt. Throughout this submittal Cycle I was used as a reference for Cycle 2.

The staff hat eviewed the submitted information and the supporting documents regardingfus49esign,nucleardesign, thermal-hydraulicdesign, plant transients and accident analysis. Our evaluation follows.

2.0 EVALUATION 2.1 Fuel. Mechanical. Design gj No changes in the fuel mechanical design basis have occurred in the fabrication of the Batch D fuel. Some design changes were made to improve fuel handling and burnup capabilities of the poison rods. These changes involve: (1) the upper end fitting hold down plate to improve handling, (b) a fuel assembly inspection envelope is changed to a square of 8.290 inches on the side for the e

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A entire assembly length, (c) the poison rod assembly design was modified to replace the solid Zircaloy-4 spacers with hollow Zircaloy-4 tubes, and (d) replacement of the two piece lower end fitting center post with a one piece casting.

The staff has found Reference 3 acceptable where clad collapse analyses are not necessary for new Combustion Engineering manufactured fuel because of the absence of gaps between pellets.

We find the above changes to be minor improvements which do not affect the mechanical design basis and, thus, are acceptable. ,

2.2 Thermal Design The Cycle 2 thermal performance evaluation was based on the performance of a I composite fuel pin that envelopes the pin performance of all of the fuel '

present in Cycle 2, i.e., Batches B, C, and D. The evaluation was performed using the NRC approved code FATES 3A (Refs. 4-7) and a power history enveloping the power and burnup levels representative of the peak pin at each burnup interval from the beginning of cycle to the ed of burpup (Ref.4). The peak pin burnup analyzed is in excess of that expected at the end of Cycle 2. Based :)

on this analysis, the internal pressure in the most limiting fuel rod will be !

1,146.8 psia which is far below the reactor coolant pressure of 2,250 psia. '

This satisfies the SRP requirements and is acceptable.

- 2.3 Nuclear Design 2.3.1 Fuel Management The Cycle 2 core will consist of 73 Batch B assemblies, 64 Batch C and 104 Batch D !

(new) assemblies. The Cycle 2 loading is low leakage, using previously burned-assemblies in the periphery. Thus, most of the Batch D assemblies are located throughout the core interior. The expected Cycle 2 lifetime is 410 effective ,

t full power days. The highest Batch D enrichment is 3.9 w/o U-235 which is l lower than the 4.05 w/o U-235 for which the Palo Verde facilities have been i approved for fuel storage. Comparison of characteristic physics parameter for l Cycle 2 and the reference cycle shows that the two cycles vary little from each other. ;

2.3.2 Po M 91stribution Calculated all-rods-out relative assembly power densities were provided for the beginning, middle and end of cycle. Relative assembly power densities for rodded configurations were also presented. The rodded configurations are those allowed by the power dependent insertion limit at full power. The nominal axial peaking factors are estimated to range from 1.14 to 1.11 at the beginning and end of Cycle 2, respectively. Augmentation factors have been eliminated ,

from this cycle as discussed in Reference 8. The methodology for the physics 1 L and power distribution calculations is based on ROCS-DIT (with the MC module) which has beer, approved by the NRC (Refs. 9,10). These calculations based on approved methods are acceptable.

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2.3.3 Control. Requirements The most restrictive value of the shutdown margin occurs at the end of cycle under hot zero power conditions. The minimum shutdown margin required to control the reactivity transient resulting from a steam line break is 6.5%

del ta-k/ k. This shutdown margin is assured as discussed in paragraph 2.5.3.

In addition sufficient boration capability and control element assembly worth with a stock control element assembly exist to meet these shutdown requirements.

These results were derived with approved methods and incorporate conservative assumptions; therefore, the results are acceptable.

2.4 Thermal-Hydraulic Design Steady state thermal-hydraulic analyses for Cycle 2 were performed using the approved code TORC (Ref. 10), the Combustion Engineering CE-1 critical heat flux correlation (Ref. 11) and the CETOP code described in Reference 12. The methodologies described in References 10-12 with the statistical combination of uncertainties (Ref. 13) the core protection system, the core operating limit system and the DNBR value of 1.24 assure- that at the 95/95 confidence / probability level the hot rod will not experience DNB. The 1.24 value includes all applicable penalties, such as the rod bow for burnups to 30,000 MWD /MTU, the .01 DNBR for the HID-1 grids and the penalties specified in the statistical combination of uncertainties (Ref.14-16). The rod bow value used in the analysis is 1.75% DNBR, foi varnups up to 30,000 MWD /MTU. For burnups higher than 30,000 MWD /MTV suffitient margin exists to offset the rod bow penalty due to lower radial power peaks in these higher burnup assemblies and rods, hence, the rod bow penalty is adequate for all anticipated burnups.

l We conclude that the thermal-hydraulic design analyses were performed using

approved codes and accounted for all applicable penalties, and therefore, are I acceptable.

2.5 Safety. Analyses.(Non-LOCA)

The design basis events considered in this safety analysis are classified in l two groups: The anticipated operational occurrences (moderate frequency and infrequent evouts) and the limiting fault events i.e., postulated accidents.

All events avaluated with respect to four criteria: fuel performance

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(centerline , reactor coolant system pressure, loss of shutdown margin and offsite dose. All events were reevaluated to assure that they meet their respective criteria for Cycle 2. The limiting events for each criterion and l those not bounded by the Cycle 1 values were reanalyzed. The analytical methodology for the reanalyses are the same as for Palo Verde Unit 3 Cycle 1 1.e., the reference cycle as described in the FSAR, the CESSAR and Ref.17.

All of the methodologies used have been reviewed and approved by the NRC. The 1

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I following list includes the code, the purpose for which was used in the j analyses and the reference: )

i Purpose Code Ref.

CESEC-III Plant response to non-LOCA events 17 CETOP-D Hot channel and DNBR 12 ]

TORC Pin DNBR and RCP shaft seizure 10,18 '

.CENPD-183 Loss-of-flow methodology analysis 19 HERMITE Core simulation for space-time kinetics 20 The input parameters for the analyses were comparable.to those for the !

reference cycle. Whenever the core protection system trip was evoked in the sequence the instrument channel response times assumed were conservative relative to the applicable Technical Specifications.

All of the events evaluated are bounded by the reference cycle except:

control element assembly misoperation, asymmetric steam generator events and 4 steam line break. Each of these limiting events analysis is discussed below. l 2.5.1 Control Element Assembly Misoperation A control element assembly (CEA) misoperation is defined as the inadvertent release of a single CEA (or a CEA subgroup) causing it to drop into the core.

The single full length and part length control element assemblies were !

reanalyzed to determine the initial thermal margin that must be maintained by the limiting conditions of operation such that the DNBR and the full centerline melt limit will not be violated. Because the control assembly downward position penalty factors have been eliminated, a 4-fingered control assembly drop will not generate a trip, therefore, sufficient margin must be maintained by the limiting conditions of' operation. However, for the i 12-fingered control assemblies the core protection calculator will provide a trip. The method used to analyze the single control assembly drop event is described in Reference 21.

The single full length CEA was analyzed because this event requires the i maximum initi . margin to be maintained by the limiting conditions of j operation. initial conditions were selected conservatively, for example.

the turbine . is not reduced, causing a power mismatch, the cycle most negative moder~etor and fuel temperature coefficients were assumed, the charging pumps and pressurizer heaters are assumed inoperable (to maximize i pressure drop) and all other systems are assumed to be on manual mode having no effect on the transient. The event is initiated by dropping a full length CEA over 1.0 sec. The largest power peaking was obtained by examining the configurations allowed by the power dependent insertion limit, which resulted in a peaking factor increase of 8.5% and a minimum DNBR greater than 1.24 at ]

15 minutes into the event. However, before this time i.e., at 10 minutes the j operator will take actitd, ?9 reduce power if the CEA has not been realigned, '

according to Fig. 3.1-2A of the Technical Specification. A maximum allowable

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initial heat generation rate of 18.0 Kw/ft could exist as an~ initial condition ,

without' exceeding the acceptable steady state fuel centerline' melt'11mit of-21'.0 Kw/ft. The. heat rate for this transient is-based on the more limiting LOCA. initial heat. rate of 13.5 Kw/ft. 'j 1'

Considering the'above we conclude that the control assembly misoperation analysis meets the requirements of the SRP Section 15.4.3 and is acceptable. ]

2.5.2 Assymmetric. Steam. Generator Events- '

Of the four events which could affect a steam generator it has been determined that the loss of load to a single steam generator is the most limiting' event.

. Such total loss of load could result if both~ main steam isolation valves were inadvertently closed. This sequence of. events causes an. initial' increase of

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temperature and-pressure and decrease of the water level of the-affected steam generator. Pressure will increase, which might cause the secondary safety valves to open. In addition the steam generator low level trip may be-activated. On the primary side the increased water temperature will result in '

a temperature tilt across the core with power increases on the cold side which could potentially approach the DNBR and the fuel melt limits. The core protection system, has a high cold leg differential temperature trip which is the primary means of protection.

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This event is. analyzed using the CESEC code for the NSSS response (Ref.17).

The resulting case parameters form the input to the HERMITE code, to model the ,

effects of the space-time radial power tilt (Ref. 20). Finally the thermal.

margin changes are evaluated using the CETOP code (Ref. 12). 1 The results. indicated that the minimum DNBR is greater than 1.24, based upon a high differential' cold ~1eg trip being generated at 6.0 second into'the event.

We find'the methods and the results acceptable.

2.5.3. Steam Line Break A steam line break can take place inside or outside containment. Inside containment may cause environmental degradation to cansor inputs and/or the core protection system. However, the variable overpower trip, which includes input from the s$sistance temperature detectors and the excore detectors, can be assumed f 6 a1 under all conditions.

The case which'was found to be less conservative with respect to the reference cycle is the steam line break at zero power, because the cooldown reactivity insertion curve is more adverse than the reference cycle curve. In addition, a sweepout volume of 119 fta before safety injection reaches the RCS was

- assumed for the post-trip return to power case as opposed to the 34.7 ft3 accounted for in the reference cycle. The effect of this reactivity was accommodated in Cycle 2 by increasing the shutdown margin required by the Technical Specifications at zero power from 6% delta-k/k to 6.5% delta-k/k. :

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.' i Considering the above results we find the steam line break analysis acceptable.

2.6 ECCS Analyses An ECCS analysis was performed for the limiting break size LOCA for Cycle 2 to demonstrate compliance with the requirements of 10 CFR 50.46. The methodology is the same as for the Cycle 1 analysis (Ref. 22). The analysis justifies a 13.5 Kw/f t peak linear heat generation rate. For Cycle 2, since there have been no significant changes in hardware characteristics, only clad temperatures and oxidation are required in this reevaluation. The code STRIKIN-Il was used for this purpose (Ref. 23). The performance data were generated with the FATES-3A fuel evaluation code (Refs. 5 & 6). It was demonstrated that the double ended guillotine break with a discharge coefficient of 1.0 is the limiting size. Similarly the limiting burnup, i.e., with the highest fuel stored energy, was found to be at 1,000 MWD /MTU. The CCCS analysis methods discussed above have been previously approved and are acceptable.

2.6.1 Large LOCA Analysis The input data compared to the reference cycle were conservative. The results for the limiting double ended guillotine break showed a peak clad temperature of 1,957'F, peak clad oxidation, 5.8% and total core-wide oxidation less than

.80%. All these values are within the required 10 CFR 50.46' limits of 2,200 F, 17.0% and 1.0% respectively. Therefore, we find the large LOCA analysis results to be acceptable.

2.6.2 Small Break.LOCA. Analysis Review of the Cycle 2 fuel and core data confirmed that the small break LOCA dnalysis' results are bounded by the corresponding results of the reference cycle.

3.0 TECHNICAL SPECIFICATION CHANGES j l

This section provides a summary of the proposed amendments to the Palo Verde !

Unit 3 Technir4 Specifications for the Cycle 2 operation. A brief' description, justification %5d acceptability for each Technical Specification (TS) change is provided in ( following.

TS 3.1.1.2: 'The proposed change raises the required shutdown margin for hot l zero power conditions from 6.0% delta-k/k to 6.5% delta-k/k to accommodate the j requirements for the steam line break. This change is necessary to satisfy regulatory requirements and thus, is acceptable. l TS Fig 3.1-1: The main change is a broadening of the moderator temperature j coefficient (MTC) range. Another is the change of the abscissa from average 4 moderator temperature to core power. The increased MTC range is necessary from ;

the increased enrichment of the new fuel. The maximum positive MTC at low j 5

-power is 5.0 pcm/*F, (1 pcm=10" delta k/k). This value satisfies accident j l

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g analyses requirements, is within the limits of the ATWS requirements (Ref.16) and satisfies GDC 11, therefore, this change is necessary and acceptable. The change of the abscissa is a matter of convenience, is not substantive and is acceptable.

TS 3.2.8: The proposed change affects the operational pressure band for the pressurizer pressure, from 1,815-2,370 psia to 2,025-2,300 psia. The reason is to support the core protection calculator improvement program. The proposed change narrows the permissible band and is, therefore, conservative ano acceptable.

TS.3.1.3.1, 3.1.3.2, 3.10.2 and.3.10.4: The proposed change separates the part length control element assemblies (PLCEA) from TS 3.1.3.1 and 3.1.3.2 to a separate TS 3.1.3.7. In' addition the special test exception for the PLCEA are modified to reference the new TS 3.1.3.7. The proposed change constitutes an improvement over the existing TSs, because it is an explicit reference to the PLCEAs, allowing to differentiate their operational requirements. Therefore, these changes are accepteble.

TS 3.3.1 Table 3.3-2: The proposed amendment changes the response time of the DNBR low reactor coolant pump shaft speed trip in the Technical Specification Table 3.3-2. The change from 0.70 sec to 0.30 sec response time was necessitated by the redefinition of the trip condition. The shaft speed trip redefinition is proposed to eliminate unnecessary spurious trips of the system. We find this change an improvement and, therefore, acceptable.

TS 3/4.1.3.5 and 3/4.1.3.6.and Fig. 3.1-3 and.3.1-4: The proposed amendment constitutes a revision of the ex1 sting T5 to accress control element assembly (CEA) insertion limits with one or two CEAs out of service. This revision ris necessitated by the change in the CEA worth in the Cycle 2 core physics. The revisions are_ necessary to maintain shutdown margin, therefore, are j acceptable. 1 i TS Tables 3.3-2.and.3.3 2a: The proposed amendment eliminates core f protection calculator penalties which compensate for the resistance l temperature detector response times greater than 8 sec. The Cycle 2 safety analyses assume a maximum resistance temperature detector response time of 8 sec and do notHaclude calculator penalty factors for response times greater than 8 sec. T W removal of the penalty factors is acceptable because these factors are ncE% sed in the Cycle 2 safety analyses.

TS.2.1.1.1,. Table.2.2 1, Bases.2.1.1.and.2.2.1*: The proposed amendment changes references of the calculated departure from nucleate boiling ratio (DNBR) from 1.231 to 1.24. This amendment also deletes references to the calculation of additional rod bow penalties if the penalty incorporated into the DNBR limit is not sufficient for any part of the cycle.

l l The methodology and the results discussed in section 2.5.2 yielded a DNBR limit of 1.24. This value continues to ensure that power operation limits

This specification change should include DNBR change to 1.24.in the Bases of 3/4.4.1 which was not included in the licensee submittal request.

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8- 1 calculations are conservative. This is acceptable. As for the rod bow penalty, it is now more advantageous to use a large enough DNBR penalty to provide protection throughout the cycle, rather than to apply incremental penalty f actors. A rod bow penalty factor of 1.75'; DNBR provides sufficient margin for burnups up to 30,000 MWD /MTV. For greater burnups, there is sufficient margin from other factors to offset any small increase in the roo .;

bow penalty. '

TS 3.2.5: The proposed amendment changes the6 reactor coolant system (RCS) total t1gw from greater or eoual to 164.0x10 lbm/hr to greater or equal to 155.8x10 lbm/hr. This change is consistent with the value of flow used in the plant safety analyses and is therefore acceptable. The wording of the specification has also been modified to clarify the need to account for instrument error in the comparison of measured flow to the safety analysis ,

value. This clarification is an editorial enhancement and is, therefore, !

acceptable.

1 TS 3/4.2.1: The proposed amendment changes the linear heat rate (LHR) limit from 14.0 Kw/ft to 13.5 Kw/ft. This change was the result of the safety i reanalysis for Cycle 2 in order to ensure that peak clad temperature does not exceed the 10 CFR 50.46 limits. In addition the amendment delineates how the LHR is to be monitored. This TS change is based on acceptable safety analysis and, therefore, is acceptable.

TS 3/4.2.4, Table 3.3 1,. Bases 3/4.2.4.and.3/4.1.3: There are a number of '

I proposed amendments to ensure reactor operation within approved safety limits, increase operator reliability and increase clarity. The proposed changes in each TS are discussed below.

Section 3.2.4 is replaced by a new format which addresses the specific {'

conditions supervisoryfor monitoring)DNBR system (COLSS and/orwith the or without control the core element operating assembly limit clusters. (

These cover the following four cases: (a)COLSSisoperableandeither !

or both CEACs are operable (b) COLSS operable but neither CEAC is operable (c) COLSS inoperable and one or both CEACs are operable and (d)

COLSS out of service and neither CEAC is operable. The new format states the action statements for all four of the above cases and replaces Figures 3.2-1 andi3'.2-2 with new Figures 3.2-1, 3.2-2 and 3.2-2a.

Section 35f-1-This section implements the following changes introduced in 3/4.2.4 above i.e., (a) removes references to reactor operation witn or without COLSS operable and both CEACs inoperable and (b) deletes the graph of DNBR operating limit margin (fig. 3.3-1) based on COLSS for both CEACs inoperable. (This is a result of this information having been incorporated into TS 3/4.2.4)

Section 3/4.1.3 Bases, changes to reflect the changes in 3/4.2.4 above.

Section 3.2.4 Bases, change to reflect changes in 3.2.4.

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The above changes ensure operation of Cycle 2 within the approved safety analyses and increase the specifications' wording clarity. Therefore, they -

are acceptable.

Figure 3.2-4a This amendment modifies the azimuthal tilt operating limits with the core operating limit supervisory system in operation, to avoid.

lengthy delays in increasing power. When the core operating limit supervisory ;

system is in operation, reactor operation within the analysis limits is assured, therefore, the proposed amendment is acceptable.

TS 3.3.2, Table 3.3-4: The proposed modification removes the " greater than" sign from the trip value of the refueling water storage tank actuation signal in Table 3.3-4. This removes an ambiguity concerning the level setpoint and is acceptable.

TSs 3/4.3.1, 3/4.3.2 and 2.2.1, Bases: The proposed changes constitute administrative changes to the bases of 3.3.1, 3.3.2 and 2.2.1 to ensure clarity and conciseness, to include updated references and to remove Cycle 1 information no longer applicable to Cycle 2. These changes are, therefore, acceptable.

4.0 STARTUP TESTING The licensee presented a description of the planned startup testing, which includes: low power physics, ascension to power and procedures if acceptance j criteria are not met. The objective of the testing is to verify that the core i performance is consistent with the design and safety analyses. The program lements the i conforms to the requirements normal surveillance requirementsof the ANSI of the /ANS-19.6.1, Technical 1985 and supp(Refs. 24 &

Specifications 25). The lower power physics tests include: initial criticality, critical boron concentration, temperature reactivity coefficient, control element assembly reactivity worth and inverse boron worth. The power ascension testing includes: flux symmetry verification, core power distribution, shape annealing matrix, boundary point power correlation coefficient, radial peaking factors, control element assembly shadowing factor, reactivity coefficient at power and. critical boron concentration. These tests will provide reasonable ,

assurance that t,he core has been loaded in accordance with the safety analysis assumptions. They are therefore acceptable.

Should any ofM startup tests reveal any unreviewed safety issues the NRC will be notiffed.

5.0

SUMMARY

AND CONCLUSIONS

! The Reactor Systems Branch reviewed the submitted information in support of the Palo Verde Unit 3 Cycle 2 operation. The review covered fuels, physics, thermal hydraulics, accident and transient analyses, technical specification revisions and startup test procedures.

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Based on the evaluations presented in the preceding sections we find the i proposed reload acceptable. !

6.0 CONTACT WITH STATE OFFICIAL The Arizona Radiation Regulatory Agency was advised of the proposed determi-nation of no significant hazards consideration with regard to these changes.

No comments were received.

7.0 ENVIRONMENTAL CONSIDERATION

S This amendment involves a change in the installation or use of facility com- ;

ponents located within the restricted area as defined in 10 CFR 20 relating to a reactor refueling. The staff has determined that this amendment involves r:, ;

significant increase in the amount, and no significant change in the type, of i any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Comission has previously issued proposed findings that the amendment involves no significant hazard consideration, and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment. -

8.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safet be endangered by operation in the proposed manner, (2) activities such y of the public will bewill not ;

conducted in compliance with the Commission's regulations, and (3) the-issuance ;

of this amendment will not be inimical to the common defense and security or to I the health and safety of the public. We therefore, conclude that the proposed. j changes are acceptable.

Principal Contributor: L. Lois !

Dated: June.9. 1989 I

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9.0 REFERENCES

1. . Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 3, Proposed Reload Technical Specification Changes," dated December 14, 1988.

2. Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "Palo Verde Nuclear Generating Station Unit 3, Submittal of the Reload Analysis Report," dated December 27, 1988.

3. Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC " Generic Applicability of EPRI NP-3966-CCM, Volume 5"(161-01867-JRP/DBK) dated April 26, 1989.

4 Letter from D.B. Karner, Arizona Nuclear Power Project to USNRC, "

Operating History of the Reference Cycle" (161-01866-08K/JRP) dated April 26, 1989.

5. CENPD-139-P-A, '!C-E Fuel Evaluation Model," Combustion Engineering, dated July 1974.

6. CEN-161(B)-P, " Improvements in the Fuel Evaluation.Model," Combustion Engineering, dated July 1981.

7. Letter from R. A. Clark (NRC) to A.fi. Lundvall, Jr. (BG&E), " Safety EvaluationofCEN-161(FATES 3),"datedMarch 31, 1983.

8. CENPD-153P, Rev.1-P-A, " INCA /CECOR Power Peaking Uncertainty,"

Combustion Engineering, dated May 1980.

9. CENPD-266-PA, "The ROCS and DIT Computer Codes for Nuclear Design,"

Combustion Engineering, dated April 1983.

10. CENPD-161-PA, " TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," Combustion Engineering, dated April 1986.

11. CENPD-162-A, " Critical Heat Flux Correlation for C-E fuel Assemblies with Standard $~

Combustig$. pacer Grids,

, Engineering, dated Part 1, Uniform September 1976. Axial Power Distribution,"

12. CEN-160-5Y'Rev.1-P, "CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Unit 2 and 3," Combustion Engineering, dated September 1981.

13. CEN-356-V-PA, Rev. 1-PA, " Modified Statistical Combination of Uncertainties," Combustion Engineering, dated May 1988.

14. CENPD-225-PA, " Fuel and Poison Rod Bowing," Combustion Engineering, dated June 1983.

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15. Letter from A.E. Scherer Combustion Engineering to'D.G. Eisenhut NRC (Enclosure 1), " Statistical Combination .of System Parameter' Uncertainties-in Thermal Margin Analyses for System 80," dated May 14, 1982. <

,. - l 4

3 16. CESSAR SSER 2 Section 4.4.6, " Statistical Combination of Uncertainties,"

Combustion Engineering, {j 1

17. 'CESEC, " Digital Simulation of a Combustion Engineering Nuclear Steam. 'l Supply System," Combustion Engineering enclosure 1-P to LD-82-001, dated January 6,1982.

18. CENPD-206-P, " TORC Code Verification and Simplified Modeling Methods,"

Combustion Engineering, dated January 1977.

CENPD-183, " Loss of Flow, CE Method for Loss-of-Flow Analysis,"

Combustion Engineering, dated July 1975.

20. CENPD-188-A, "HERMITE, Space Time Kinetics," Combustion Engineering, dated July 1975.

21. CENPD-199-PA, Rev. IP, "CE Setpoint Methodology," Combustion Engineering, dated January 1986.

22. CENPD-132-P, " Calculative Methods for the CE Large Break LOCA Evaluation Model," Combustion Engineering, dated August 1974. Also Supplements 1 and 2--dated December 1974. and July 1975 respectively.

23. CENPD-135-P, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer a Program," Combustion Engineer .g, dated April 1974. Also Supplements 2P and 4P dated February 1975 and August 1975 respectively.

24. ANSI /ANS-19.6.1-1985, " Reload Startup Physics Tests for' Pressurized Water Reactors."

25. CEN-319 " Control Rod Group Exchange Technique," Combustion Engineering, dated November 1985.

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v t e Letter Sequence Approval
TAC:71574, (Approved, Closed)
Initiation Acceptance... Administration Questions Answers Results Approval Other:
MONTHYEAR1989-06-23 [Table View]

ML20245J224

Text

Dear Mr. Conway:

SUBJECT:

Enclosures:

Dear Mr. Conway:

SUBJECT:

Enclosures:

Enclosure:

1.0 INTRODUCTION

SUMMARY

7.0 ENVIRONMENTAL CONSIDERATION

8.0 CONCLUSION

9.0 REFERENCES

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