REG 3- UNITED STATES NUCLEAR REGULATORY COMMISSION · Akin, Gump, Strauss, Hauer and Feld El Paso Electric Company 1333 New Hampshire Avenue, Suite 400 Washington, DC. 20036 Ms. Angela

- � REG

4/

0

UNITED STATES

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

September 30, 1994

Mr. William L. Stewart Executive Vice President, Nuclear Arizona Public Service Company Post Office Box 53999 Phoenix, Arizona 85072-3999

SUBJECT: ISSUANCE OF AMENDMENTS FOR THE PALO VERDE NUCLEAR GENERATING STATION UNIT NO. I (TAC NO. M88992), UNIT NO. 2 (TAC NO. M88991), AND UNIT NO. 3 (TAC NO. M88993)

Dear Mr. Stewart:

The Commission has issued the enclosed Amendment No.82 to Facility Operating License No. NPF-41, Amendment No.69 to Facility Operating License No. NPF-51, and Amendment No.54 to Facility Operating License No. NPF-74 for the Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3, respectively. The amendments consist of changes to the Technical Specifications in response to your application dated February 18, 1994, as supplemented by your letter dated June 20, 1994.

These amendments allow credit to be taken for burnup of spent fuel assemblies in establishing storage locations within the spent fuel storage pool. The current spent fuel storage pool is configured to store fresh fuel assemblies with a maximum radially average enrichment of 4.30 weight percent (w/o) U-235 in a two-out-of-four checkerboard array. These amendments allow for three distinct storage regions. Region 1 allows storage of fresh fuel assemblies with a maximum radially averaged enrichment equal to 4.30 w/o U-235 in a checkerboard configuration. Region 2 allows storage of spent fuel assemblies in a three-out-of-four configuration. Region 3 allows storage of spent fuel assemblies in every location (four-out-of-four configuration).

9410120292 940930 PDR ADOCK 05000528 P PDR

.Z ? - 3-

Mr. William L. Stewart

A copy of the related Safety Evaluation is also enclosed. issuance will be included in the Commission's next regular Register notice.

A notice of biweekly Federal

Sincerely,

Original signed by:

Brian E. Holian, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Docket Nos. STN 50-528, STN 50-529 and STN 50-530

Enclosures: 1 . 2. 3. 4.

cc w/enclosures: See next page

DISTRIBUTION

EAd en mDHagan, T4A43 OPA, 02G5 JRoe TQuay CGrimes, 011E22 Region IV Liran ADummer

Amendment No. 82 to NPF-41 Amendment No. 69 to NPF-51 Amendment No. 54 to NPF-74 Safety Evaluation

NRC & Local PDRs DFoster-Curseen GHill (6), T5C3 OC/LFDCB, T9EIO PDIV-2/RF OGC, 015B18 ACRS (10), TWFN BHolian KPerkins, WCFO LKopp

nflnrIMFNT NAMF. PvR~qq? -AMfl *See Previous Sheet for Concurrence_____ PW P04-2/PM PD4-2/PA ~ SRXB*

NAME ] Foster-Curseen LTranLU BHolian:pk RJones

DATE___ //-/94 C1 1/1 9 4 j I/o/94 9/15/94

NAME CMcCracken LCunninghafi MlA(' TQuay

DATE 19/15/94 1 9/21/94 k/I\ __4 4_I__b/19 4

OFC S PLB*

OFFICIAL RECUKU CUOY

PRPB* OGC I PD4-2/D

F' �

� 48

September 30, 1994-2 -

&#

Mr. William L. Stewart

A copy of the related Safety Evaluation is also enclosed. issuance will be included in the Commission's next regular Register notice.

A notice of biweekly Federal

Sincerely,

Brian E. Holian, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Docket Nos. STN 50-528, STN 50-529 and STN 50-530

Enclosures: I . 2. 3. 4.

Amendment No.82 to Amendment No.69 to Amendment No.54 to Safety Evaluation

cc w/enclosures: See next page

NPF-41 NPF-51 NPF-74

-2 -

Mr. William L. Stewart Arizona Public Service Company

cc: Mr. Steve Olea Arizona Corporation Commission 1200 W. Washington Street Phoenix, Arizona 85007

T. E. Oubre, Esq. Southern California Edison Company P. 0. Box 800 Rosemead, California 91770

Senior Resident Inspector Palo Verde Nuclear Generating Station 5951 S. Wintersburg Road Tonopah, Arizona 85354-7537

Regional Administrator, Region IV U. S. Nuclear Regulatory Commission Harris Tower & Pavillion 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064

Mr. Charles B. Brinkman, Manager Washington Nuclear Operations ABB Combustion Engineering Nuclear Power 12300 Twinbrook Parkway, Suite 330 Rockville, Maryland 20852

Mr. Aubrey V. Godwin, Director Arizona Radiation Regulatory Agency 4814 South 40 Street Phoenix, Arizona 85040

Palo Verde

Jack R. Newman, Esq. Newman & Holtzinger, P.C. 1615 L Street, N.W., Suite Washington, D.C. 20036

1000

Mr. Curtis Hoskins Executive Vice President and

Chief Operating Officer Palo Verde Services 2025 N. 3rd Street, Suite 220 Phoenix, Arizona 85004

Roy P. Lessey, Jr., Esq. Akin, Gump, Strauss, Hauer and Feld El Paso Electric Company 1333 New Hampshire Avenue, Suite 400 Washington, DC. 20036

Ms. Angela K. Krainik, Manager Nuclear Licensing Arizona Public Service Company P. 0. Box 52034 Phoenix, Arizona 85072-2034

Chairman, Maricopa County Board of Supervisors

111 South Third Avenue Phoenix, Arizona 85003

UNITED STATES

I' • NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

ARIZONA PUBLIC SERVICE COMPANY, ET AL.

DOCKET NO. STN 50-528

PALO VERDE NUCLEAR GENERATING STATION, UNIT NO. 1

AMENDMENT TO FACILITY OPERATING LICENSE

Amendment No. 82

License No. NPF-41

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

A. The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project

Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company

of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated February 18, 1994, as supplemented by letter dated June 20, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I;

B. The facility will operate in conformity with the application, the

provisions of the Act, and the rules and regulations of the Commission;

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

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

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

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

amended to read as follows:

9410120295 940930 PDR ADOCK 05000528 P PDR

-2-

(2) Technical Specifications and Environmental Protection Plan

The Technical Specifications contained in Appendix A, as revised through Amendment No. 8Z and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3. This license amendment is effective as of the date of issuance and must be fully implemented no later than 45 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

4\1

Theodore R. Quay, Director Project Directorate IV-2 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation

Attachment: Changes to the Technical

Specifications

Date of Issuance: September 30, 1994

ATTACHMENT TO LICENSE AMENDMENT

AMFNflMFNT NA. �2 TO FACTITTY OPERATING LICENSE NO. NPF-41


Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove

IX XIV XIX

B 3/4 9-3 5-5 5-6

Insert

IX XIV XIX 3/4 9-17 B 3/4 9-3 5-5 5-6 5-6a

AMENDMENT NO 82

INDEX

LIMITING CONDITIONS Fbt,-OPERATION AND SURVEILLANCE RE9biEMENTS

SECTION

ELECTRICAL POWER SYSTEMS (Continued)

3/4.8.2 D.C. SOURCES

OPERATING ............................................ SHUTDOWN .............................................

3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS

OPERATING ............................................ SHUTDOWN .............................................

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES .................................

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES .................................

3/4.9 REFUELING OPERATIONS

3/4.9.1 BORON CONCENTRATION .....................................

3/4.9.2 INSTRUMENTATION .........................................

3/4.9.3 DECAY TIME ..............................................

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS .......................

3/4.9.5 COMMUNICATIONS ..........................................

3/4.9.6 REFUELING MACHINE .......................................

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING .........

3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION

HIGH WATER LEVEL ..................................... LOW WATER LEVEL ......................................

3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ................

3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES ...................................... CEAs .................................................

3/4.9.11 WATER LEVEL - STORAGE POOL ..............................

3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM ..............

3/4.9.13 BORON CONCENTRATION - SPENT FUEL STORAGE POOL ...........

3/4.10 SPECIAL TEST EXCEPTIONS

3/4.10.1 SHUTDOWN MARGIN AND KN-i - CEA WORTH TESTS ..............

3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS .............

3/4.10.3 REACTOR COOLANT LOOPS ...................................

AMENDMENT NO. -3, 82

PAGE

3/4 8-9 3/4 8-13

3/4 8-14 3/4 8-16

3/4 8-17

3/4 8-40

3/4

3/4

3/4

3/4

3/4

3/4

3/4

9-1

9-2

9-3

9-4

9-5

9-6

9-7

3/4 9-8 3/4 9-9

3/4 9-10

3/4 3/4

3/4

3/4

3/4

9-11 9-12

9-13

9-14

9-17

3/4 10-1

3/4 10-2

3/4 10-3

I

IXPALO VERDE - UNIT 1

INDEX

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

SECTION PAGE 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS

AND REACTOR COOLANT COLD LEG TEMPERATURE ................ 3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY ........ 3/4 10-5 3/4.10.6 SAFETY INJECTION TANKS .................................. 3/4 10-6 3/4.10.7 SPENT FUEL POOL LEVEL ................................... 3/4 10-7 3/4.10.8 SAFETY INJECTION TANK PRESSURE .......................... 3/4 10-8 3/4.10.9 SHUTDOWN MARGIN AND KN-1 - CEDMS TESTING ................ 3/4 10-9 3/4.10.10 NATURAL CIRCULATION TESTING PROGRAM ..................... 3/4 10-10

3/4.11 RADIOACTIVE EFFLUENTS

3/4.11.1 LIQUID HOLDUP TANKS ..................................... 3/4 11-1

3/4.11.2 EXPLOSIVE GAS MIXTURE ................................... 3/4 11-2

3/4.11.3 GAS STORAGE TANKS ....................................... 3/4 11-3

AMENDMENT NO. 00, 62PALO VERDE - UNIT I X

INDEX

BASES

SECTION PAGE

3/4.7 PLANT SYSTEMS

3/4.7.1 TURBINE CYCLE ........................................... B 3/4 7-1

3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION ......... B 3/4 7-3

3/4.7.3 ESSENTIAL COOLING WATER SYSTEM .......................... B 3/4 7-3

3/4.7.4 ESSENTIAL SPRAY POND SYSTEM ............................. B 3/4 7-4

3/4.7.5 ULTIMATE HEAT SINK ...................................... B 3/4 7-4

3/4 7.6 ESSENTIAL CHILLED WATER SYSTEM .......................... B 3/4 7-4

3/4.7.7 CONTROL ROOM ESSENTIAL FILTRATION SYSTEM ................ B 3/4 7-5

3/4.7.8 ESF PUMP ROOM AIR EXHAUST CLEANUP SYSTEM ................ B 3/4 7-5

3/4.7.9 SNUBBERS ................................................ B 3/4 7-5

3/4.7.10 SEALED SOURCE CONTAMINATION ............................. B 3/4 7-7

3/4.7.11 SHUTDOWN COOLING SYSTEM ................................. B 3/4 7-7

3/4.7.12 CONTROL ROOM AIR TEMPERATURE ............................ B 3/4 7-7

3/4.8 ELECTRICAL POWER SYSTEMS

3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS ................ B 3/4 8-1

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ................. B 3/4 8-3


3/4.9.1 BORON CONCENTRATION ..................................... B 3/4 9-1

3/4.9.2 INSTRUMENTATION ......................................... B 3/4 9-1

3/4.9.3 DECAY TIME .............................................. B 3/4 9-1

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................... B 3/4 9-1

3/4.9.5 COMMUNICATIONS .......................................... B 3/4 9-1

3/4.9.6 REFUELING MACHINE ....................................... B 3/4 9-2

PALO VERDE - UNIT 1 AMENDMENT NO. 23XIII

INDEX

BASES

SECTION


3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION ................


3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL ...........................................




3/4.10.1 SHUTDOWN MARGIN AND K,.- - CEA WORTH TESTS ..............

3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS...............


3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE ................

3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY ........

3/4.10.6 SAFETY INJECTION TANKS ..................................

3/4.10.7 SPENT FUEL POOL LEVEL ...................................

3/4.10.8 SAFETY INJECTION TANK PRESSURE .......................... 3/4.10.9 SHUTDOWN MARGIN AND KN.-1 - CEDMS TESTING .................


B

B

B

B

B

B

PAGE

3/4 9-2

3/4 9-2

3/4 9-3

3/4

3/4

3/4

9-3

9-3

9-3

B 3/4 10-1

B

B

B

B

B

B

B

B

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

10-1

10-1

10-1

10-1

10-2

10-2

10-2

10-2

LIQUID HOLDUP TANKS .....................................

EXPLOSIVE GAS MIXTURE ...................................

GAS STORAGE TANKS .......................................

AMENDMENT NO. 23--362, 82

3/4.11.1

3/4.11.2

3/4.11.3

B

B

B

3/4

3/4

3/4

11-1

11-1

11-1

PALO VERDE - UNIT I XIV

INDEX

LIST OF FIGURES

3.1-1

3.2-1

3.4-1

3.4-2a

3.4-2b

3.4-2c

3.4-2d

4.7-1

5.1-1

5.1-2

5.1-3

5.6-1

AMENDMENT NO. W ,-6-9-, 82

MINIMUM BORATED WATER VOLUMES ..........................

REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVEL ..................................................

DOSE EQUIVALENT 1-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1.0 liCi/GRAM DOSE EQUIVALENT 1-131 ...................

REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR LESS THAN 8 EFPY OF OPERATION ..........

REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS FOR 8 TO 32 EFPY OF OPERATION ..............

REACTOR COOLANT SYSTEM MAXIMUM ALLOWABLE HEATUP AND COOLDOWN RATES FOR LESS THAN 8 EFPY OF OPERATION ..............................................

REACTOR COOLANT SYSTEM MAXIMUM ALLOWABLE HEATUP AND COOLDOWN RATES FOR 8 TO 32 EFPY OF OPERATION ..............................................

SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST ..............

SITE AND EXCLUSION BOUNDARIES ..........................

LOW POPULATION ZONE ....................................

GASEOUS RELEASE POINTS .................................

ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT ..............

PAGE

3/4 1-11

3/4 2-8

3/4 4-27

3/4 4-29

3/4 4-29a

3/4 4-29b

3/4 4-29c

3/4 7-26

5-2

5-3

5-4

5-6aI

PALO VERDE - UNIT 1 XIX

INDEX

LIST OF TABLES

PAGE

1.1 FREQUENCY NOTATION ...................................... 1-8 1.2 OPERATIONAL MODES ....................................... 1-9 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT

LIMITS .................................... , ............. 2-3 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION ........ , ............. 3/4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES ....... 3/4 3-11 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE

REQUIREMENTS .............................. 4 ............. 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 3-31 3.3-6 RADIATION MONITORING INSTRUMENTATION .................... 3/4 3-38 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS ............................................ 3/4 3-40 3.3-7 SEISMIC MONITORING INSTRUMENTATION ...................... 3/4 3-43 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS ............................................ 3/4 3-44 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION ............... 3/4 3-46 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION

SURVEILLANCE REQUIREMENTS ............................... 3/4 3-47 3.3-9A REMOTE SHUTDOWN INSTRUMENTATION ......................... 3/4 3-49 3.3-98 REMOTE SHUTDOWN DISCONNECT SWITCHES ..................... 3/4 3-50

AMENDMENT NO. 17 69PALO VERDE - UNIT I XX

REFUELING OPERATIONS

3/4.9.13 BORON CONCENTRATION - SPENT FUEL STORAGE POOL

LIMITING CONDITION FOR OPERATION

3.9.13 The boron concentration in the spent fuel storage pool shall be maintained greater than or equal to 2150 ppm.

APPLICABILITY: Whenever fuel assemblies are in the spent fuel storage pool.

ACTION:

With the requirements of this specification not satisfied, suspend all movement of fuel assemblies in the spent fuel storage pool and restore the boron concentration to 2150 ppm or greater. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS

4.9.13 The boron concentration of the spent fuel storage pool shall be determined by chemical analysis at least once per 7 days when fuel assemblies are in the spent fuel storage pool.

PALO VERDE UNIT 1 AMENDMENT NO. 823/4 9-17


BASES

A shutdown cooling loop may be removed from operation for up to 1 hour per 8-hour period during surveillance testing of ECCS pumps. This is necessary to meet Surveillance 4.5.2, flow testing of the HPSI pumps without other pumps running, and 4.3.3.5, testing of the containment spray pumps and LPSI pumps during surveillance of the remote shutdown system.

3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM

The OPERABILITY of this system ensures that the containment purge valves will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and STORAGE POOL

The restrictions on minimum water level ensure that sufficient water depth (at least 23 feet above the top of the spent fuel) is available to remove a nominal 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly for a maximum fuel rod pressurization of 1200 psig. The minimum water depth is consistent with the assumptions of the safety analysis.

3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM

The limitations on the fuel building essential ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.

The use of ANSI Standard N509 (1980) in lieu of ANSI Standard N509 (1976) to meet the guidance of Regulatory Guide 1.52, Revision 2, Positions C.6.a and C.6.b, has been found acceptable as documented in Revision 2 to Section 6.5.1 of the Standard Review Plan (NUREG-0800).


The restriction on the boron concentration of the spent fuel storage pool ensures that keff is maintained less than or equal to 0.95 in the event that either a new or spent fuel assembly is either improperly loaded into any available storage location in the spent fuel storage pool, or is dropped in the gap between the pool wall and a Region 3 storage rack. This boron concentration is more than adequate to ensure that the k f limit of 0.95, delineated in Specification 5.6.1.1.a, will not be violated.

AMENDMENT NO. 4-3-, 82PALO VERDE - UNIT 1 B 3/4 9-3

DESIGN FEATURES

5.3 REACTOR CORE

FUEL ASSEMBLIES

5.3.1 The reactor core shall contain 241 fuel assemblies with each fuel assembly normally containing 236 fuel rods or burnable poison rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel may be made if justified by a cycle specific reload analysis. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of approximately 1950 grams uranium. Each burnable poison rod shall have a nominal active poison length of 136 inches. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum radially averaged enrichment of 4.30 weight percent U-235 at any axial location.

CONTROL ELEMENT ASSEMBLIES

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

5.4 REACTOR COOLANT SYSTEM

DESIGN PRESSURE AND TEMPERATURE

5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable surveillance requirements,

b. For a pressure of 2500 psia, and

c. For a temperature of 650 0F, except for the pressurizer which is 7000F.

VOLUME

5.4.2 The total water and steam volume of the Reactor Coolant System is 13,900 + 300/-0 cubic feet at a nominal Tv, of 593°F.

IAMENDMENT NO. 4,-74, 82PALO VERDE - UNIT 1 5-5

DESIGN FEATURES

5.5 METEOROLOGICAL TOWER LOCATION

5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE

5.6.1 CRITICALITY

5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. The maximum calculated kef value, including margin for uncertainty in calculational method and mechanical tolerances, less than or equal to 0.95 with a 95% probability at a 95% confidence level when flooded with unborated water.

b. A nominal 9.5 inches center-to-center distance between adjacent storage cell locations.

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

5.6.1.3 The spent fuel storage pool is organized into three regions for spent fuel storage. Fuel shall be placed in the appropriate region based on appropriate initial enrichment and existing burnup as designated in Figure 5.6-1:

a. Region 1: Fuel shall be stored in a checkerboard (two-out-offour) storage pattern. Fuel that qualifies to be stored in Regions 1, 2, or 3, in accordance with Figure 5.6-1, may be stored in Region 1.

b. Region 2: Fuel shall be stored in a three-out-of-four storage pattern. Fuel that qualifies to be stored in Regions 2 or 3, in accordance with Figure 5.6-1, may be stored in Region 2.

c. Region 3: Fuel shall be stored in a four-out-of-four storage pattern. Only fuel that qualifies to be stored in Region 3, in accordance with Figure 5.6-1, shall be stored in Region 3.

DRAINAGE

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

CAPACITY

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

5 7 COMPONENT CYCLIC OR TRANSIENT LIMITS

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

PALO VERDE - UNIT 1 5-6 AMENDMENT NO. 82

FIGURE 5.6-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT

500DO

-- 45000 E

*1-40000

S35000

-3 0 0 0 0

S25000

>,,200D0 -Fo E 15000

(D

10000

5000

02.50 3.00 3.50 4.00 Initial Enrichment, wt

PALO VERDE - UNIT I AMENDMENT NO. 82

4.30

REGION 3

,REGION 2

/J REGION 1

II I'/ rr 1 1 1 6 11 l tI i= J = -rrT

4.50 5.00

cxx0 4/4 ***** 3/4

2.00

5-6a

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20&55-0001





Amendment No. 69

License No. NPF-51


A. The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated February 18, 1994, as supplemented by letter dated June 20, 1994, 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 Part I;

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




2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-51 is hereby amended to read as follows:

-2-


The Technical Specifications contained in Appendix A, as revised through Amendment No. 69, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.



Theodore R. Quay, Director Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation


Specifications

September 30, 1994Date of Issuance:


AMENDMENT NO. 69 TO FACILITY OPERATING LICENSE NO. NPF-51



Remove Insert

IX XIV XIX

B 3/4 9-3 5-5 5-6

IX XIV XIX 3/4 9-17 B 3/4 9-3 5-5 5-6 5-6a

INDEX


SECTION PAGE



OPERATING ............................................ 3/4 8-9 SHUTDOWN ........................................ 3/4 8-13


OPERATING ....................................... 3/4 8-14 SHUTDOWN ........................................ 3/4 8-16


CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ................................. 3/4 8-17

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES ................................. 3/4 8-40


3/4.9.1 BORON CONCENTRATION ..................................... 3/4 9-1

3/4.9.2 INSTRUMENTATION ......................................... 3/4 9-2

3/4.9.3 DECAY TIME .............................................. 3/4 9-3

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................... 3/4 9-4

3/4.9.5 COMMUNICATIONS .......................................... 3/4 9-5

3/4.9.6 REFUELING MACHINE ....................................... 3/4 9-6

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING ......... 3/4 9-7


HIGH WATER LEVEL ..................................... 3/4 9-8 LOW WATER LEVEL ...................................... 3/4 9-9

3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ................ 3/4 9-10

3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES .................................. 3/4 9-11 CEAs ................................................. 3/4 9-12

3/4.9.11 WATER LEVEL - STORAGE POOL .............................. 3/4 9-13

3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM .............. 3/4 9-14

3/4.9.13 BORON CONCENTRATION - SPENT FUEL STORAGE POOL ........... 3/4 9-17


3/4.10.1 SHUTDOWN MARGIN AND Kw. 1 - CEA WORTH TESTS .............. 3/4 10-1

3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ............. 3/4 10-2

3/4.10.3 REACTOR COOLANT LOOPS ................................... 3/4 10-3

AMENDMENT NO.-437, 69PALO VERDE - UNIT 2 IX

�MJ

INDEX


SECTION





3/4.10.8 SAFETY INJECTION TANK PRESSURE ..........................

3/4.10.9 SHUTDOWN MARGIN AND KN. 1 - CEDMS TESTING ................


3/4.11.1 LIQUID HOLDUP TANKS .....................................

3/4.21.2 EXPLOSIVE GAS MIXTURE ...................................

3/4.11.3 GAS STORAGE TANKS .......................................

PAGE

3/4

3/4

3/4

3/4

3/4

3/4

10-4

10-5

10-6

10-7

10-8

10-9

3/4 11-1

3/4

3/4

11-2

11-3

PALO VERDE - UNIT 2 AMENDMENT NO. 11, 48X

INDEX

BASES

SECTION PAGE

3/4.7 PLANT SYSTEMS

3/4.7.1 TURBINE CYCLE ........................................... B 3/4 7-1

3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION ......... B 3/4 7-3

3/4.7.3 ESSENTIAL COOLING WATER SYSTEM .......................... B 3/4 7-3

3/4.7.4 ESSENTIAL SPRAY POND SYSTEM ............................. B 3/4 7-4

3/4.7.5 ULTIMATE HEAT SINK ...................................... B 3/4 7-4

3/4 7.6 ESSENTIAL CHILLED WATER SYSTEM .......................... B 3/4 7-4

3/4.7.7 CONTROL ROOM ESSENTIAL FILTRATION SYSTEM ................ B 3/4 7-5

3/4.7.8 ESF PUMP ROOM AIR EXHAUST CLEANUP SYSTEM ................ B 3/4 7-5

3/4.7.9 SNUBBERS ................................................ B 3/4 7-5

3/4.7.10 SEALED SOURCE CONTAMINATION ............................. B 3/4 7-7

3/4.7.11 SHUTDOWN COOLING SYSTEM .................................. B 3/4 7-7

3/4.7.12 CONTROL ROOM AIR TEMPERATURE ............................. B 3/4 7-7


3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS ................... B 3/4 8-1

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES ................. B 3/4 8-3


3/4.9.1 BORON CONCENTRATION ..................................... B 3/4 9-1

3/4.9.2 INSTRUMENTATION ......................................... B 3/4 9-1

3/4.9.3 DECAY TIME .............................................. B 3/4 9-1

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................... B 3/4 9-1

3/4.9.5 COMMUNICATIONS .......................................... B 3/4 9-1

3/4.9.6 REFUELING MACHINE ....................................... B 3/4 9-2

PALO VERDE - UNIT 2 AMENDMENT NO. 13XIII

-- i~lxi_

INDEX

BASES

SECTION








3/4.10.1 SHUTDOWN MARGIN AND KN-1 - CEA WORTH TESTS ..............

3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ................







3/4.10.9 SHUTDOWN MARGIN AND KN-1 - CEDMS TESTING ................


B

B

B

PAGE

3/4 9-2

3/4 9-2

3/4 9-3

B 3/4 9-3

B 3/4 9-3

B 3/4 9-3

B 3/4 10-1

B 3/4

B 3/4

B 3/4

B 3/4

B 3/4

B 3/4

B 3/4

B 3/4

10-1

10-1

10-1

10-1

10-2

10-2

10-2

10-2

LIQUID HOLDUP TANKS .....................................

EXPLOSIVE GAS MIXTURE ...................................

GAS STORAGE TANKS .......................................

AMENDMENT NO. -I-3,-48, 69

3/4.11.1

3/4.11.2

3/4.11.3

B

B

B

3/4

3/4

3/4

11-1

11-1

11-1

I

PALO VERDE - UNIT 2 XlV

INDEX

LIST OF FIGURES

3.1-I

3.2-1

3.4-1

AMENDMENT NO. 4-2-, -&--, 69



DOSE EQUIVALENT 1-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1.0 pCi/GRAM DOSE EQUIVALENT 1-131 ...................










3.4-2a

3.4-2b

3.4-2c

3.4-2d

4.7-1

5.1-1

5.1-2

5.1-3

5.6-1

PAGE

3/4 1-11

3/4 2-8

3/4 4-27

3/4 4-29

3/4 4-29a

3/4 4-29b

3/4 4-29c

3/4 7-26

5-2

5-3

5-4

5-6a I


INDEX

LIST OF TABLES

PAGE

1.1 FREQUENCY NOTATION ...................................... 1-8 1.2 OPERATIONAL MODES ....................................... 1-9 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT

LIMITS .................................................. 2-3

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

3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES......... 3/4 3-11

4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................ 3/4 3-14

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

3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ............................. 3/4 3-25

3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES ............... 3/4 3-28 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM

INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............... 3/4 3-31 3.3-6 RADIATION MONITORING INSTRUMENTATION .................... 3/4 3-38 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS ............................................ 3/4 3-40 3.3-7 SEISMIC MONITORING INSTRUMENTATION ...................... 3/4 3-43 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE

REQUIREMENTS ............................................ 3/4 3-44 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION ............... 3/4 3-46 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION

SURVEILLANCE REQUIREMENTS ............................... 3/4 3-47 3.3-9A REMOTE SHUTDOWN INSTRUMENTATION ......................... 3/4 3-49

3.3-9B REMOTE SHUTDOWN DISCONNECT SWITCHES ..................... 3/4 3-50

AMENDMENT NO. z, 55

ý- - A--L-Ll-

PALO VERDE - UNIT 2 XX






ACTION:



4.9.13 The boron concentration of the spent fuel determined by chemical analysis at least once per are in the spent fuel storage pool.

storage pool shall be 7 days when fuel assemblies

PALO VERDE - UNIT 2 iAMENDMENT NO. 693/4 9-17


BASES

A shutdown cooling loop may be removed from operation for up to I hour per 8-hour period during surveillance testing of ECCS pumps. This is necessary to meet Surveillance 4.5.2, flow testing of the HPSI pumps without other pumps running, and 4.3.3.5, testing of the containment spray pumps and LPSI pumps during surveillance of the remote shutdown system.









The restriction on the boron concentration of the spent fuel storage pool ensures that keff is maintained less than or equal to 0.95 in the event that either a new or spent fuel assembly is either improperly loaded into any available storage location in the spent fuel storage pool, or is dropped in the gap between the pool wall and a Region 3 storage rack. This boron concentration is more than adequate to ensure that the k • limit of 0.95, delineated in Specification 5.6.1.1.a, will not be violated.

PALO VERDE - UNIT 2 B 3/4 9-3 AMENDMENT NO.-7;, 69

DESIGN FEATURES

5.3 REACTOR CORE

FUEL ASSEMBLIES

5.3.1 The reactor core shall contain 241 fuel assemblies with each fuel assembly normally containing 236 fuel rods or burnable poison rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel may be made if justified by a cycle specific reload analysis. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of approximately 1950 grams uranium. Each burnable poison rod shall have a nominal active poison length of 136 inches. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum radially averaged enrichment of 4.30 weight percent U-235 at any axial location.








c. For a temperature of 650 0 F, except for the pressurizer which is 7000F.

VOLUME

5.4.2 The total water and steam volume of the Reactor Coolant System is 13,900 + 300/-0 cubic feet at a nominal Tavg of 593 0F.

IAMENDMENT NO.-l 1 --- , 69PALO VERDE - UNIT 2 5-5

DESIGN FEATURES



5.6 FUEL STORAGE

5.6.1 CRITICALITY


a. The maximum calculated keff value, including margin for uncertainty in calculational method and mechanical tolerances, less than or equal to 0.95 with a 95% probability at a 95% confidence level when flooded with unborated water.







DRAINAGE


CAPACITY


5 7 COMPONENT CYCLIC OR TRANSIENT LIMITS


PALO VERDE - UNIT 2 AMENDMENT NO. 695-6

FIGURE 5.6-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT-

50000

.- '45000

E "ý'40000

-'35000

a30000

j 25000

>,,20000

-0

E 15000

(F) 10000

5000

(. Ivpi9i iv9 v. i v -. .. . . . . . .. . . . ..... 4o I--I I I I v I ............ 1

2.50 3.00 3.50

Initial Enrichment,

PALO VERDE - UNIT 2 5-6a AMENDMENT NO. 69

4.30

REGION 3

SREGION 2 /_

REGION 1

__/-

4.00 4.50

wt5.00

0O 4/4 L* -***-* 3/4

I

2.00

UNITED STATES

0 NUCLEAR REGULATORY COMMISSION 0 ; •WASHINGTON, D.C. 20555-0001





Amendment No. 54 License No. NPF-74


A. The application for amendment by the Arizona Public Service Company (APS or the licensee) on behalf of itself and the Salt River Project Agricultural Improvement and Power District, El Paso Electric Company, Southern California Edison Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority dated February 18, 1994, as supplemented by letter dated June 20, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I;

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




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

-2-


The Technical Specifications contained in Appendix A, as revised through Amendment No.54 , and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. APS shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.



Theodore R. Quay, Director Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation


Specifications

Date of Issuance: September 30, 1994


£IMFNflMFNT NA '�A TO FACILITY OPERATING LICENSE NO. NPF-74



Remove

IX XIV XIX

B 3/4 9-3 5-5 5-6

Insert

IX XIV XIX 3/4 9-17 B 3/4 9-3 5-5 5-6 5-6a

TO FACILITY OPERATING LICENSE NO. NPF-74AmmnFN NnIU 9dII 1 •

INDEX

LIMITING CONDITIONS Fru<OPERATION AND SURVEILLANCE REOQiREMENTS

SECTION PAGE



OPERATING ............................................ 3/4 8-9 SHUTDOWN ............................................. 3/4 8-13


OPERATING ............................................ 3/4 8-14 SHUTDOWN ............................................. 3/4 8-16


CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES ................................. 3/4 8-17

MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES ................................. 3/4 8-40


3/4.9.1 BORON CONCENTRATION ..................................... 3/4 9-1

3/4.9.2 INSTRUMENTATION ......................................... 3/4 9-2

3/4.9.3 DECAY TIME .............................................. 3/4 9-3

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS ....................... 3/4 9-4

3/4.9.5 COMMUNICATIONS .......................................... 3/4 9-5

3/4.9.6 REFUELING MACHINE ....................................... 3/4 9-6

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING ......... 3/4 9-7


HIGH WATER LEVEL ..................................... 3/4 9-8 LOW WATER LEVEL ...................................... 3/4 9-9

3/4.9.9 CONTAINMENT PURGE VALVE ISOLATION SYSTEM ................ 3/4 9-10

3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES ...................................... 3/4 9-11 CEAs ................................................. 3/4 9-12

3/4.9.11 WATER LEVEL - STORAGE POOL .............................. 3/4 9-13

3/4.9.12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM .............. 3/4 9-14

3/4.9.13 BORON CONCENTRATION - SPENT FUEL STORAGE POOL ........... 3/4 9-17


3/4.10.1 SHUTDOWN MARGIN AND K,-. - CEA WORTH TESTS .............. 3/4 10-1

3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ............. 3/4 10-2

3/4.10.3 REACTOR COOLANT LOOPS ................................... 3/4 10-3

AMENDMENT NO.--2-, 54PALO VERDE - UNIT 3 IX

INDEX


SECTION PAGE 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS

AND REACTOR COOLANT COLD LEG TEMPERATURE ................ 3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY ........ 3/4 10-5 3/4.10.6 SAFETY INJECTION TANKS .................................. 3/4 10-6 3/4.10.7 SPENT FUEL POOL LEVEL ................................... 3/4 10-7 3/4.10.8 SAFETY INJECTION TANK PRESSURE .......................... 3/4 10-8 3/4.10.9 SHUTDOWN MARGIN AND KN.1 - CEDMS TESTING ................ 3/4 10-9 3/4.10.10 NATURAL CIRCULATION TESTING PROGRAM ..................... 3/4 10-10


3/4.11.1 LIQUID HOLDUP TANKS ..................................... 3/4 11-1

3/4.11.2 EXPLOSIVE GAS MIXTURE ................................... 3/4 11-2

3/4.11.3 GAS STORAGE TANKS ....................................... 3/4 11-3

PALO VERDE - UNIT 3 X AMENDMENT NO. 2, 34

INDEX

BASES

SECTION

3/4.7 PLANT SYSTEMS

3/4.7.1 TURBINE CYCLE ...........................................

3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION .........

3/4.7.3 ESSENTIAL COOLING WATER SYSTEM ..........................

3/4.7.4 ESSENTIAL SPRAY POND SYSTEM ............................

3/4.7.5 ULTIMATE HEAT SINK ......................................

3/4 7.6 ESSENTIAL CHILLED WATER SYSTEM......................

3/4.7.7 CONTROL ROOM ESSENTIAL FILTRATION SYSTEM ................

3/4.7.8 ESF PUMP ROOM AIR EXHAUST. CLEANUP SYSTEM ................

3/4.7.9 SNUBBERS ................................................

3/4.7.10 SEALED SOURCE CONTAMINATION .............................

3/4.7.11 SHUTDOWN COOLING SYSTEM .................................

3/4.7.12 CONTROL ROOM AIR TEMPERATURE ............................


3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS ................

3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES .................


3/4.9.1 BORON CONCENTRATION .....................................

3/4.9.2 INSTRUMENTATION .........................................

3/4.9.3 DECAY TIME .................. ; ...........................

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS .......................

3/4.9.5 COMMUNICATIONS ..........................................

3/4.9.6 REFUELING MACHINE .......................................

PAGE

B

B

B

B

B

B

B

B

B

B

B

B

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

7-1

7-3

7-3

7-4

7-4

7-4

7-5

7-5

7-5

7-7

7-7

7-7

B 3/4 8-1

B 3/4 8-3

B

B

B

B

B

B

3/4

3/4

3/4

3/4

3/4

3/4

9-1

9-1

9-1

9-1

9-2

PALO VERDE - UNIT 3

I

XIII AMENDMENT NO. 2

INDEX

BASES

SECTION








3/4.10.1

3/4.10.2

3/4.10.3

3/4.10.4

3/4.10.5

3/4.10.6

3/4.10.7

SHUTDOWN MARGIN AND KN. 1 - CEA WORTH TESTS ...............

MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS ................

REACTOR COOLANT LOOPS ...................................

CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE ................

MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY ........

SAFETY INJECTION TANKS ..................................

SPENT FUEL POOL LEVEL ...................................


3/4.10.9 SHUTDOWN MARGIN AND K N-i - CEDMS TESTING .................


3/4.11.1 LIQUID HOLDUP TANKS .....................................

3/4.11.2 EXPLOSIVE GAS MIXTURE ...................................

3/4.11.3 GAS STORAGE TANKS .......................................

B 3/4 10-1

B

B

B

B

B

B

B

B

B

B

B

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

3/4

10-1

10-1

10-1

10-1

10-2

10-2

10-2

10-2

11-1

11-1

11-1

AMENDMENT NO. -, -34, 54

PAGE

3/4 9-2

3/4 9-2

3/4 9-3

B

B

B

B

B

B

3/4

3/4

3/4

9-3

9-3

9-3

PALO VERDE - UNIT 3 XlV

INDEX

LIST OF FIGURES

3.1-1

3.2-1

3.4-1

3.4-2a

3.4-2b

3.4-2c

3.4-2d

4.7-1

5.1-1

5.1-2

5.1-3

5.6-1



DOSE EQUIVALENT 1-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1.0 uCi/GRAM DOSE EQUIVALENT 1-131 ...................










AMENDMENT NO. .28, -42-, 54

PAG.E

3/4 1-11

3/4 2-8

3/4 4-27

3/4 4-29

3/4 4-29a

3/4 4-29b

3/4 4-29c

3/4 7-26

5-2

5-3

5-4

5-6a

LIST OF

FIGURES

I


INDEX

LIST OF TABLES

PAGE

1.1 FREQUENCY NOTATION ...................................... 1-8

1.2 OPERATIONAL MODES ....................................... 1-9

2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS ................................................ 2-3

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

3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES ....... 3/4 3-11

4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................ 3/4 3-14

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

3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ............................. 3/4 3-25

3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES ............... 3/4 3-28

4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............... 3/4 3-31

3.3-6 RADIATION MONITORING INSTRUMENTATION .................... 3/4 3-38

4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................ 3/4 3-40

3.3-7 SEISMIC MONITORING INSTRUMENTATION ...................... 3/4 3-43

4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................................ 3/4 3-44

3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION ............... 3/4 3-46

4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ............................... 3/4 3-47

3.3-9A REMOTE SHUTDOWN INSTRUMENTATION ......................... 3/4 3-49

3.3-98 REMOTE SHUTDOWN DISCONNECT SWITCHES ..................... 3/4 3-50

AMENDMENT NO. iS, 42PALO VERDE - UNIT 3 XX






ACTION:



4.9.13 The boron concentration of the spent fuel storage pool shall be determined by chemical analysis at least once per 7 days when fuel assemblies are in the spent fuel storage pool.

PALO VERDE - UNIT 3 I3/4 9-17 AMENDMENT NO. 54


BASES

A shutdown cooling loop may be removed from operation for up to 1 hour per 8-hour period during surveillance testing of ECCS pumps. This is necessary to meet Surveillance 4.5.2, flow testing of the HPSI pumps without other pumps running, and 4.3.3.5, testing of the containment spray pumps and LPSI pumps during surveillance of the remote shutdown system.









The restriction on the boron concentration of the spent fuel storage pool ensures that keff is maintained less than or equal to 0.95 in the event that either a new or spent fuel assembly is either improperly loaded into any available storage location in the spent fuel storage pool, or is dropped in the gap between the pool wall and a Region 3 storage rack. This boron concentration is more than adequate to ensure that the k f limit of 0.95, delineated in Specification 5.6.1.1.a, will not be violated.

PALO VERDE - UNIT 3 AMENDMENT NO. 54B 3/4 9-3

DESIGN FEATURES

5.3 REACTOR CORE

FUEL ASSEMBLIES

5.3.1 The reactor core shall contain 241 fuel assemblies with each fuel assembly normally containing 236 fuel rods or burnable poison rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4 or stainless steel may be made if justified by a cycle specific reload analysis. Substitution of up to a total of 80 fuel rods clad with zirconium-based alloys other than Zircaloy-4 may also be made in two fuel assemblies for in-reactor performance evaluation purposes during Cycles 4, 5 and 6. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of approximately 1950 grams uranium. Each burnable poison rod shall have a nominal active poison length of 136 inches. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum radially averaged enrichment of 4.30 weight percent U-235 at any axial location.








c. For a temperature of 650 0F, except for the pressurizer which is 7000F.

VOLUME

5.4.2 The total water and steam volume of the Reactor Coolant System is 13,900 + 300/-0 cubic feet at a nominal T,, of 593°F.

AMENDMENT NO. &--45, 46-, 54PALO VERDE - UNIT 3 5-5

DESIGN FEATURES



5.6 FUEL STORAGE

5.6.1 CRITICALITY


a. The maximum calculated keff value, including margin for uncertainty in calculational method and mechanical tolerances, less than or equal to 0.95 with a 95% probability at a 95% confidence level when flooded with unborated water.







DRAINAGE


CAPACITY


5.7 COMPONENT CYCLIC OR TRANSIENT LIMITS


PALO VERDE - UNIT 3 AMENDMENT NO. 545-6

FIGURE 5.6-1 ASSEMBLY BURNUP VERSUS INITIAL ENRICHMENT.

50000

-- '-45000

E •'•40000

S35000

0- 3 0 0 0 0 C D 25000

>,20000 -o E 15000

U) 1)i0000)

5000

0 2..1l)U z.Du .5.u .50 4.UU

Initial Enrichment, wt4.5o 5.00

4/4 3/4

PALO VERDE - UNIT 3

4.30

REGION 3

REGION 2

S~REGION 1

tI--rTrr

GGQGW

5-6a AMENDMENT NO. 54

* • , REG1j

UNITED STATES So NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO AMENDMENT NO. 82 TO FACILITY OPERATING LICENSE NO. NPF-41.

AMENDMENT NO. 69 TO FACILITY OPERATING LICENSE NO. NPF-51,

AND AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NO. NPF-74


PALO VERDE NUCLEAR GENERATING STATION, UNIT NOS. 1. 2. AND 3

DOCKET NOS. STN 50-528, STN 50-529, AND STN 50-530

1.0 INTRODUCTION

By letter dated February 18, 1994, the Arizona Public Service Company (APS or the licensee) submitted a request for changes to the Technical Specifications (TS) for the Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3 (Appendix A to Facility Operating License Nos. NPF-41, NPF-51, and NPF-74, respectively). The Arizona Public Service Company submitted this request on behalf of itself, the Salt River Project Agricultural Improvement and Power District, Southern California Edison Company, El Paso Electric Company, Public Service Company of New Mexico, Los Angeles Department of Water and Power, and Southern California Public Power Authority. The proposed changes would allow credit to be taken for burnup of spent fuel assemblies in establishing storage locations within the spent fuel storage pool.

The current spent fuel storage pool is configured to store fresh fuel assemblies with a maximum radially average enrichment of 4.30 weight percent (w/o) U-235 in a two-out-of-four checkerboard array. The proposed changes would allow for three distinct storage regions. Region 1 would allow storage of fresh fuel assemblies with a maximum radially averaged enrichment equal to 4.30 w/o U-235 in a checkerboard configuration. Region 2 would allow storage of spent fuel assemblies in a three-out-of-four configuration. Region 3 would allow storage of spent fuel assemblies in every location (four-out-of-four configuration). Allowable storage in Region 2 or 3 depends upon the initial assembly enrichment and the assembly burnup, as shown in TS Figure 5.6-1. In each fueled location, a stainless steel L-shaped insert is used to position the fuel and maintain the minimum edge-to-edge spacing between assemblies.

The licensee supplemented their original amendment with a letter dated June 20, 1994. This letter provided responses to a staff request for additional information. The information was of a clarifying nature and did not affect the staff's original no significant hazards determination.

9410120296 940930 PDR ADOCK 05000528 P PDR

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2.0 EVALUATION

The analysis of the reactivity effects of fuel storage in the spent fuel storage racks was performed by ABB-Combustion Engineering (CE) using the twodimensional discrete ordinates transport theory DOT-IV computer code, with four energy group neutron cross sections generated by the CEPAK code. These codes have been previously used by CE for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments. These experiments simulate the PVNGS fuel storage racks as realistically as possible with respect to parameters important to reactivity such as enrichment and assembly spacing. In March 1992, the NRC issued Information Notice 92-21 and Supplement I concerning discrepancies that were discovered in spent fuel pool reactivity calculations. The discrepancies were due to an overestimation of neutron absorption in the CEPAK generation of cross sections. These discrepancies were found to exist only in regions containing a strong neutron absorber (poison). Since neutron poison is not present, this problem does not exist for the PVNGS racks. The staff concludes that the analytical methods used are acceptable and capable of predicting the reactivity of the PVNGS storage racks with a high degree of confidence.

Region I will be comprised of three 9x8 storage racks, one 12x8 storage rack, and one 9x9 storage rack. To prevent inadvertent storage of a fuel assembly in a cell required to be vacant, the cell blocking devices currently in place in every other storage cell location will remain to maintain a two-out-of-four checkerboard configuration. Therefore, the configuration of Region I is identical with the current spent fuel pool configuration, and the previously approved criticality analysis remains applicable.

Region 2 will be comprised of three 9x8 storage racks and one 12x8 storage rack. Since storage in Region 2 will be limited to a three-out-of-four storage arrangement, cell blocking devices will be employed in one out of every four storage cell locations to preclude the possibility of an unanalyzed assembly configuration. Region 3 will be comprised of six 9x8 storage racks and two 12x8 storage racks. Since fuel assemblies may be stored in every Region 3 cell location, no cell blocking devices will be installed in Region 3. Cell blocking devices will also be placed along the Region 2 interface with Region 3 to eliminate the possibility of an unanalyzed arrangement of assemblies.

The modeling of Regions 2 and 3 included several conservative assumptions. These assumptions neglected the reactivity effects of axial leakage, poison shims in the assemblies, structural grids, and soluble boron in the 68 'F pool water. These assumptions tend to increase the calculated effective multiplication factor (keff) of the racks and are, therefore, acceptable. The stored fuel assemblies were modeled as CE 16x16 assemblies with a nominal pitch of 0.506 inches between fuel rods, a fuel pellet diameter of 0.33 inches, and a U02 density of 10.4 g/cc.

DOT-IV calculations were used to construct a curve of burnup versus initial enrichment for both Regions 2 and 3 (TS Figure 5.6-1) such that all points on

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the curve produce a keff value (without uncertainties or biases) of 0.93. This method of reactivity equivalencing has been accepted by the NRC and used for numerous other spent fuel storage pools which take credit for burnup. The NRC criticality acceptance criterion for fuel storage is that k ff be no greater than 0.95, including all uncertainties at a 95% probability/ 95% confidence level. Therefore, the reactivity effects due to uncertainties in minimum center-to-center pitch, eccentric positioning of assemblies, minimum monolith thickness, temperature variations, minimum L-insert thickness, assembly enrichment, and assembly burnup were obtained as well as a methodology uncertainty and bias. These were applied to the nominal value of 0.93 to obtain a final k ff of 0.944 for the spent fuel racks. This meets the NRC criterion of no greater than 0.95 and is, therefore, acceptable.

Most abnormal storage conditions will not result in an increase in the keff of the racks. However, it is possible to postulate events, such as an assembly drop on top of a rack or between a rack and the pool walls or the misloading of an assembly, with a burnup and enrichment combination outside of the acceptable area in TS Figure 5.6-1, which could lead to an increase in reactivity. However, for such events, credit may be taken for the presence of 2150 ppm of boron in the pool water required by TS 3.9.13, since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (double contingency principle). The reduction in keff caused by the boron more than offsets the reactivity addition caused by credible accidents. Therefore, the staff criterion of keff no greater than 0.95 for any postulated accident is met.

The following TS changes have been proposed as a result of the requested amendment. The staff finds these changes acceptable.

(1) TS 3.9.13 and its associated Bases are added to require a minimum boron concentration of 2150 ppm in the spent fuel storage pool with a 7-day surveillance interval whenever fuel assemblies are in the pool. This is acceptable since this boron concentration more than offsets the reactivity addition caused by any credible accident and the staff considers a 7-day surveillance interval to be a reasonable interval to verify boron concentration.

(2) TS 5.3.1 is modified to delete the footnote requiring that "no fuel with an enrichment greater than 4.0 weight percent U-235 shall be stored in a high density mode in the spent fuel storage facility." This is acceptable since the criticality analysis crediting fuel assembly burnup has shown that initial assembly enrichments up to 4.30 w/o can be stored in the three-region pool, depending on the associated assembly burnup.

(3) TS 5.6.1 is modified to clarify the uncertainties margin as the 95% probability/95% confidence level, to establish a nominal 9.5-inch center-to-center distance between adjacent storage cell locations, to define the three regions of the spent fuel pool, and to incorporate TS Figure 5.6-1. This is acceptable since the criticality analysis has shown that the NRC acceptance criterion of keff no greater than 0.95 is met for these conditions.

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Based on the review described above, the staff finds that the criticality aspects of the proposed changes to the PVNGS spent fuel pool TS are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.

Although the PVNGS TS have been modified to specify the above-mentioned fuel as acceptable for storage in the fresh or spent fuel racks, evaluations of reload core designs (using any enrichment) will, of course, be performed on a cycle-by-cycle basis as part of the reload safety evaluation process. Each reload design is evaluated to confirm that the cycle core design adheres to the limits that exist in the accident analyses and TS to ensure that reactor operation is acceptable.

Additionally, the staff reviewed the proposed change for heat load considerations. The licensee's heat load calculations were based on a full pool with 1300 fuel assemblies, and were found to be acceptable in the NRC Safety Evaluation Report. The maximum number of fuel assemblies that can be stored in the proposed three-region configuration is 1054 fuel assemblies. The actual loading pattern will therefore have a lower decay heat than assumed in the calculations for a full pool. The heat load calculations are bounding for the three-region configuration and are, therefore, acceptable.

The staff also questioned the licensee regarding the potential adverse effects to the consequences of the fuel handling accident outside containment (this TS amendment allows a closer spacing of fuel assemblies, in certain regions of the spent fuel pool, than was allowed previously - thereby potentially allowing more fuel assemblies to be damaged on a fuel handling accident). The licensee stated that the original licensing basis allowed for spent fuel to be loaded in either a 4x4 array or a checkerboard array, depending on the use of boraflex poison. Therefore, a fuel handling accident was assumed to occur with maximum loading of the pool. Section 15.7.4 of the UFSAR, "Radiological Consequences of Fuel Handling Accidents," states:

The fuel assemblies are stored within the spent fuel rack at the bottom of the spent fuel pool. The top of the rack extends above the tops of the stored fuel assemblies. A dropped fuel assembly could not strike more than one fuel assembly in the storage rack. Impact could occur only between the ends of the involved fuel assemblies, with the lower end fitting of the dropped fuel assembly impacting against the upper end fitting of the stored fuel assembly.... Horizontal impact of a fuel assembly could result from a dropped fuel assembly falling in the horizontal position .... The worst case bundle impact results from the horizontal drop .... As a result of the fuel assembly drop, no more than four rows of fuel rods would fail.

The staff reviewed the UFSAR to ensure that the new configuration was bounded by the current analysis. Neither the individual fuel assembly characteristics nor the rack construction are changed by this TS amendment, only the location of assemblies within the spent fuel pool. The fuel pool rack construction precludes more than one assembly from being impacted in a fuel handling

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accident. Therefore, the UFSAR analysis conclusion regarding the worst scenario for a dropped assembly (in which the horizontal impact of a fuel assembly on top of the spent fuel assembly damages fuel rods in the dropped assembly but does not impact fuel in the stored assemblies) continues to be limiting, and is, therefore, acceptable.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Arizona State official was notified of the proposed issuance of the amendments. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (59 FR 17593). Accordingly, the amendments meet 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 need be prepared in connection with the issuance of the amendments.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: L. Kopp A. Dummer B. Holian

Date: September 30, 1994

Documents

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