ML20069K867

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Amends 76 & 43 to Licenses DPR-51 & NPF-6,respectively, Increasing Spent Fuel Pool Storage Capacity from 589 to 968 Locations for Unit 1 & from 485 to 988 Locations for Unit 2
ML20069K867
Person / Time
Site: Arkansas Nuclear  Entergy icon.png
Issue date: 04/15/1983
From: Clark R, Stolz J
Office of Nuclear Reactor Regulation
To:
Arkansas Power & Light Co
Shared Package
ML20069K870 List:
References
DPR-51-A-076, NPF-06-A-043 NUDOCS 8304270196
Download: ML20069K867 (27)
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UNITED STATES y "

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NUCLEAR REGULATORY COMMISSION j

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  • ARKANSAS POWER & LIGHT COMPANY DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendnent No. 76 License No. DPR-51 1.

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

A. The application. for amendment by Arkansas Power and tight Company (the licensee) dated November 5,1982, as supplemented February 17, 1983, and April 7,1983, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

Theie is reasonable assurance (1) 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 note 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 CPR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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

Technical Specifications The Technical 5pecifications contained in Appendices A and B, as revised through Amendment No. 76, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the'.

Technical Specifications.

I 3.

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

FOR THE NUCLEAR REGULATORY COMMISSION'

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hn. F.:Stolz, Chief' perating Reactors Branc #4 sio'n of Licensing '-

Attachment:

Changes to the Technical Specifications Date of Issuance: April 15,1983

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ATTACHMENT TO LICENSE AMENDMENT NO. 76 FACILITY OPERATING LICENSE NO. DPR-51 DOCKET NO. 50-313 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.

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59 59a 59b 59c (new page) 59d (new page) 116 127 f

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1.5.2-2E P.00 FOSITION LIMITS FOR TWO-PUMP OPERATION FRSM 0 TO 60 EFPD-AND-1, CYCLE 5 48c4 3.5.2-2F ROD POSITION LIMITS FOR TWO-PUMP OPERATION FROM 50 TO 200 +.

10 EFPD-ANO-1, CYCLE 5 48c5

3. 5.2-2G ROD POSITION LIMITS FOR TWO-PUMP OPERATION FROM 200 + 10 TO 400 1 10 EFPD-AND-1, CYCLE 5 48c6~

3.5.2-2H R0D POSITION LIMITS FOR TWO-PUMP OPERATION FROM 400 + 10 TO 435 1 10 EFPD-AND-1. CYCLE 5 48c7

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3.5.2-3A:

OPERATIONALPOWERIMBALANCEENVELOPEFOROPERdTIONFROM0TO 60 EFPD-ANO-1, CYCLE 5 48d 3.5.2-38 OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 50 TO 1

200 1 10 EFPD-ANO-1, CYCLE 5 48d1 l

h 3.5.2-3C OPERATIONAL POWER IMBALANCE ENVELOPE FOR OPERATION FROM 200 '+

10 TO 400 1 10 EFPD-ANO-1, CYCLE 5 48d2 3.5.2-3D OPERATIONAL POWER IMBALANCE' ENVELOPE FOR OPERATION FROM 400 +

10 TO 435 1 10 EFPD-ANO-1, CYCLE 5 48d3 1

3.5.2-4 LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE 48e 5.2-4A ASPR POSITION LIMITS FOR OPERATION FROM 0 TO 60 EFPD-ANO-1, CYCLE 5 48f 3.5.2-4B ASPR POSITION LIMITS FOR OPERATION FROM 50 TO ?.00 + 10 EFPD-AND-1, CYCLE $~~ '

48g 3.5.2-4C APSR POSITION LIMITS FOR OPERATION FROM 200 + 10 TO 400 +

10 EFPD-ANO-1, CYCLE 5 43h f

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3.5.2-4C APSR POSITION LIMITS FOR OPERATION FROM 400.+ 10 TO 435 +

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10 EFPD-ANO-1,- CYCLE 5 481 3.5-4-1 INCORE INSTRUMENTATION SPECIFICATION AXIAL IMBALANCE INDICATION 53a 3.5.4-2 INCORE INSTRUMENTATION SPECIFICATION RADIAL FLUX TILT INDICATION 53b 3.5.4-3 INCORE INSTRUMENTATION SPECIFICATION 53c 3.8.1 SPENT FUEL POOL ARRANGEMENT UNIT NO 1 59c '

3.8.2 MINIMUM BURNUP vs. INITIAL ENRICHMENT FOR REGION"2 5fd

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6.2 1 MANAG{MENTORGANIZATI0NCHART 119 6.2-2 FUNCTIONAL ORGANIZATION FOR PLANT OPERATION 120.

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AmendmentNo.g,76 y

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1 3.~.5 During tne nandling of irraciatec fuel in the reac ct builoing, at least one door on the personnel and emergency hatches shall be closed. The equipment hatch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces.

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3.8.7 Isolation valves in lines containing automatic containment

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isolation valves shall be operable, or at least one shall be closed.

i 3.8.8 When two irradiated fuel assemblies are being moved simultaneously.

i by the bridges within the fuel transfer canal, a minimum of 10 feet separation shall be maintained between the assemblies at all times.

3.8.9 If any of the above specified limiting conditions for fuel loading and refueling are not met, movement of fuel into the reactor core i

shall cease; action sha'1 be initiated to correct the conditions so that the specified limits are met, and no operations which may increase the reactivity of the core shall be made. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.10 The.raactor building purge isolation system, including the radiation monitors shall be tested and verif.ied to be operable within'7 days prior to refueling operations. The provisions of Specifications 3.D.3 and 3.0.4 are not applicable.

3.8.11

. Irradiated fuel shall not be removed from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

In the event of a complete core offload, a full core to be discharged shall be subcritical a minimum of 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> prior to discharge of more than 70 assemblies to the spent fuel. pool.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

I 3.8.12 All fuel handling in the Auxiliary Building shall cease upon notification of the issuance of a. tornado watch for Pope,- Yell, Johnson, or Logan counties in Arkansas.

Fuel handling operations in progress will be completed to the extent necessary to place the fuel handling bridge and crane in their normal parked and-locked position.

The provisions of Specifications 3'.0.3 and 3.0.4 are not applicable.

3.8.13 No loaded spent fuel shipping cask shall be carried above or into the Auxil.iary Building equipment shaft unless atmospheric dispersion conditions are equal to or better than those produced by Pasquill Type D stability accompanied by a wind velocity of 2 m/sec.

In addition, the railroad spur door of the Turbine Building shall be closed and the fuel handling area ventilation system shall be in. operation.

The provisions of Specifications i

3.0.3'and 3.0.4 are not applicable.

3.8.14 Loads in excess of 2000 pounds shall be prohibited from travel I

over fuel-asse'mblies in the storage pool.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Amendment No.-Jg,-)/,g,[ 76' 59 n

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3.8.15 The spen: fuel snipoing cask shall not ce carriec by the Auxiliary Building crane pending the evaluation of the spent fuel cask drop accident and the crane design by AP&L and NRC review and approval.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.16 Storage in the spent fuel pool shall be restricted to fuel assemblies having initial enrichment less than or equal to 4.1 w/o U-235. The provisions of Specifications 3.0.3 and 3.0.4 are not :

applicable.

3.8.17 Storage in Region 2 (as shown on Figure 3.8.1) of the spent fuel pool shall be further restricted by burnup and enrichment limits specified in Figure 3.8.2.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

3.8.18' The boron concentration in the spent fuel pool saall be maintained-(at all times) at greater than 1600 parts per million.

BASES Detailed written procedures will.be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as ddscribed in Section 9.6 of the FSAR l

incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety.

If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation.

Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.

The requirement that at least one decay heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel at the refueling temperature (normally 140'F), and (2) sufficient coolant circulation is maintainedthroughthereactorcoretominimizethegfectsofaboron dilution incident and prevent boron stratification.

The requirement to have two decay heat removal loops operable when there is less than 23 feet of water above the core,sensureg t. hat a single failure of the operating decay heat removal loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water"above the core, a large heat sink is available for core cooling, thus in the event _ of a failure of the operating decay heat removal

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loop, adequate time is provided to initiate emergency procedures to cool the Core.

The shutdown margin indicated in Specification 3.8.4 will keep tgcore-subcritical, even with all contror rods withdrawn from the c*are Although the ' refueling boron concentration is sufficient to maintain the core k 5 0.99 if all the control rods were removed from the core, only a fewco8No1rodsvillberemovedatanyonetimeduringfuelshufflingand

-Ame'ndment No I7,256, 57, 76 59a

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re:iacemen:.

The k, -itn all reds in ne core anc wi n refueling coren concentration is ap, 6ximately 0.9.

Specification 3.8.5 allows the control p

room operator to inform the reactor building personnel of any impending unsafe condition detected from the main control board indicators during fuel movement.

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l The specificatioa requiring testing reactor building purge termination is to verify that these components will function as required should a fuel handling accident occur which resulted in the release of significant fission products.

Because of physical dimensions of the fuel bridges, it is physically impossible for fuel assemblies to be within 10 feet of each other while being handled.

Specification 3.8.11 is required as:

1) the safety analysis for the fuel handling accident was gajed on the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
and, 2) to assure that the maximum design heat load of the spent fuel pool cooling system will not be exceeded during a full core offload.

Specification 3.8.14 will assure,that damage to fuel in the spent fuel pool will not be caused by dropping heavy objects onto the fuel.

Administrative controls will prohibit the storage of fuel in locations' adjoining the walls at the north and south ends of the pool, in the vicinity of cask storage area and fuel tilt pool access gates, until the review specified in 3.8.15 is completed.

Specification 3.8.15 assures that the spent fuel cask drop accident cannot l

occur prior to completion of the NRC staff's review of this potential accident and the completion of any modifications that may be necessary to preclude the accident or mitigate the consequences.

Upon satisfactory completion of the NRC's review, Specification 3.8.15 shall be deleted.

l Specifications 3.8.16 and 3.8.17 assure fuel enrichment and fuel'burnup limits assumed in the spent fuel safety anal ses will not be exceeded.

Specification 3.8.18 assures the boron concertrat' ion in the spent fuel pool i

will remain within the limits of the spent fuel pool accident and criticality analyses.

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REFERENCES (1)

FSAR, Section 9u5 (2)

FSAR, Section 14.2.2.3 (3)

FSAR, Section 14.2.2.3.3 Ame'ndment Ney-56, 57, 76 59b

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FIGURE 3,8,2 MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION 2 ENRICHMENT l\\

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Amendment No. 7g 59d

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NE'n AfC SFENT F'UEL STGF. AGE FACILIT*Ei Acolicability A:olies to storage facilities for new and spent fuel assemblies.

Objective To assure that both new and spent fuel assemblies will be stored in such a manner that an inadvertent criticality could not occ'ur.

Soecification 5.4.1 New Fuel Storage 1.

Fuel assemblies are stored in racks of parallel rows, having l

a nominal center to center distance of 21 inches in both directions.

This spacing is sufficient to maintain a K of less than.9 even if flooded with unborated water, base 8 $n f

fuel with an enrichment of 3.5 weight percent U235.

2.

New fuel may be stored in the spent fuel pool or in its-shipping containers.

5.4.2 Soent Fuel Storage 1.

The spent fuel racks are designed and shall be maintained so that the calculated effective multiplication factor is no.

greater than 0.95 (including all known uncertainties) when the pool is flooded with unborated water.

2.

The spent fuel pool and the new fuel pool racks are designed as seismic Class I equipment.

REFERENCES FSAR, Section 9.6 s

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Amendment No.17, 76 116 S

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The Nuclear Regulatory 'Cwmipion stp1 be notified end%

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3. g 6.8.1 Written procedures shall be established,s implesanted and maintained covering the activities referenced below:

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The applicable procedures racerwended in Appendix "A"Cof-Regulatory Guide 1.33,, November,1972.

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Refueling operations.

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Surveillance and test activities of safety relaQd equipment.%

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Security Plan implementation.

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Emergency Plan implementation.

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Fire Protection Program implementation. '

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New and spent fuel storage.

Y 6.8.2 Each procedure of 6.8.1 above, and changes thereto,,shall be reviewed by the PSC and approved by the General Manager prior tr implementation and reviend periodically as set forth in

  • administrative procedures.

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s 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:

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The intent of the original procedure.is not alte' ed.

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The change is approved by two members of the plant staff, at s'

least one of whom holds a Senior Reactor Operator's Li, cense s

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The change is documented, reviewed by the PSC and approved by

' the General Manager within 14" days of implementation.

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....a ARKANSAS POWER & LIGHT COMPANY DOCKET NO. 50 368 ARKANSAS NUCLEAR ONE, UNIT NO.2 AMENDMENT TO FACILITY OPERATING LICENSE

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i 3Q i i Amendment No. 43 License No..NPF-5 1.

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

A.

The application for ifnindnient by Ar. kansas fower and ;Libht C6pipany ~ ~ ~ ~ ~- ~

(the licensee) dated November 5,1982, as supplemented

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p February 17, 1983, and April 7,1983, complies with the d '\\,

standards and requirements of the Atomic Energy Act of 1954, e

. as amended (the Act), and the Commission's rules and regulations

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set forth in 10 CFR Chapter I; 8.

The, faciitty will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; s

f.A C.

There is reasonable assurance (1) 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

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conducted in compliance with the Comission's regulations; D.

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

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

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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-6 is hereby amended to read as follows:

Technical Specifications The Technical 5pecifications contained in Appendices A and B as revised through Amendment No.43, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the' Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION ~

_n obert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing -

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Attachment:

Changes to the Technical Specifications Date of Issuance: April 15,1983 9

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ATTACHMENT TO LICENSE AMENDMENT NO. 43 FACILITY OPERATING LICENSE NO. NPF-6 DOCKET NO. 50-368 Replace the following pages of the Appendix A Technical Specifications with the snclosed-pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Corresponding overleaf pages are provided to maintain document complete-ness.

Pages VIII 3/4 9-3 3/4 9-14 3/4 9-15 3/4 9-16 6-13 B 3/4 9-1 4

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INDEX 1

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves......................

3/4 7-1 Emergency Feedwater System...........................

3/4 7-5 Condensate Storage Tank..............................

3/4 7-7 Activity............................................

3/4 7-8 Main Steam' Isolation Valves.........................

3/4 7-10 1

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION...... 3/4 7-14

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3/4.7.3 SERVICE WATER SYSTEM................................. '3/4 7-15 3/4.7.4 EMERGENCY COOLING P0ND................................

3/4 7-16 3/4.7.5 FLOOD PR0TICTION.....................................

3/4 7-16a 3/4.7.6 CONTROL ROOM' EMERGENCY AIR CONDITIONING AND AIR FILTRATION SYSTEM....'..............................

3/4 7-17 3/4.7.8 HYDRAULIC SHOCK SUPPRESSORS..........................

3/4 7-22 3/4.7.9 SEALED SOURCE CONTAMINATION..........................

3/4 7-27 3.4.7.10 FIRE SUPPRESSION SYSTEMS Fire Suppression Water System........................

3/4 7-29 Spray and/or Sprinkler Systems.......................

3/4 7-33 Fire Hose Stations...................................

3/4 7-35 3/4.7.'11 PENETRATION FIRE BARRIERS............................

3/4 7-37 3/4.7.12 SPENT FUEL POOL STRUCTURAL, INTEGRITY.................

3/4 7-38 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating..............................................

3/4 8-1 Shutdown...............................'...............

3/4 8-5 ARKANSAS - UNIT 2

,VII knendment No.' 30 l___.

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'LIMITitlG CONDIT10f!S FOR OPERATIO:. AND SURVEILLA::CE REQUIREMENTS SECTION PAGE

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3/4.8.2 0NSITE POWER DISTRIBUTION SYSTEMS A. C. Distribution -

Operating............................

3/4 8-6 A. C. 'Di s tri bu tion - Shu tdown............................ 3/4 8-7 D. C. Di s tri bu tion - Opera ti ng........................... 3/4 S-8 D. C. Di s tri bu tion - Shutdown............................ 3/4 8-10 Containment Penetration Conductor Overcurrent Protective Devices..................................... 3/4 8-11 3/4.9 REFUELING.0PERATIONS 1

3/4.9.1 BORON CONC ENTRATION...................................... 3/4 9-1 3/4.9.2 INSTRUMENTATION...........................................

3/4 9-2 3/4.9.3 DECAY TIME AND SPENT FUEL ST0 RAGE......................... 3/4 9-3 l

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3/4.9.4 CONTAINMENT BUILDING PENETRATIONS.................~...'....'.

3/4 9-4 3/4.9.5 C0fiMUNICATIONS..................'......................... 3/4 9-6 3/4.9.6

~ REFUELING tiACHINE OPERABILITY............................ 3/4 9-7 3/4.9.7 CRANE TRAVEL - SPENT FUEL POOL BUILDING.................. 3/4 9-8 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION................. 3/4 9-9 3/4.9.9 WATER LEVEL - REACTOR VESSEL............................. 3/4 9-10 3/4.9.10 SPENT FUEL POOL WATER LEVEL.............................. 3/4 9-11 3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM.................... 3/4 9-12 3/4.9.12 FU EL STO RA G E.............................................

3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 S HUTDOWN MARGI N.......................................... 3 /4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS.... 3/4 10-2

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3/4.10.3 REACTOR COOLANT LOOPS.................................... 3/4 10-3 3/4.10.4 CENTER CEA MISALIGNMENT.............,.'.'................... 3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE FOR CRITICALITY...............,........

3/4 10-5 ARKANSAS - UNIT 2 VIII Amendment No. 29,43 i.--e e

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t REFUELIllG OPERATIONS DECAYTIMhAftDSPEf1TFUELSTORAGE LIMITING CONDITION FOR OPERATION 3.9.3.a The reactor shall be subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

I 3.9.3.b In the event of a complete core offload, a full core to be discharged shall be subcritical a minimum of 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> prior to discharge of more than 70 assemblies to the spent fuel pool.

APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, suspend all operations involving movement of irradiated fuel in the r,eactor pressure vessel. With the reactor subcritical for less than 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br />, suspend all operations involving movement of more than 70 fuel assemblies from the reactor-pressure

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vessel to the spent fuel pool. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.3.a The reactor shall be determined to have been suberitical for at least l

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> by verification of the date and time of subcriticality prior to move-ment of irradiated fuel in the reactor pressure vessel.

4.9.3.b The reactor shall be determined to have been subcritical for at least 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> by verification of the date and time of subcriticality prior to move-ment of the 71st irradiated fuel assembly from the reactor pressure vessel to the spent fuel pool.

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ARKANSAS - UNIT 2 3/4 9-3 Amendment No_. 43 b W e ao. e e.**

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REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a.

The equipment door closed and held in place by a minimum of four bolts, b.

A minimum of one door in each airlock is closed, and c.

Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1.

Closed by an isolation valve, blind flange, or manual valve, or 2.

Exhausting through OPERABLE containment purge -and exhaust

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, system HEPA ffiters and charcoal adsorbers.

APPLICABILITY: Duri within the containme_ng_ CORE ALTERATIONS or movement of irradiated fuel nt.

ACTION:

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With the requirements of the above specification not satisfied, immedi-ately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment. The provisions of. Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment penetrations shall be determined to be in its above required condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment.

4.9.4.2 The containment purge and exhaust system shall be demonstrated

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OPERABLE at the following frequencies:

a.

At least once per 18 months or (11 after any structural main-tenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire or chemical release in any venti-1ation zone communicating with the system by:

i ARKANSAS - UNIT 2 3/4 9-4 1

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REFUELIllG 0PERATIONS SURVEILLANCE REQUIREMEsTS *(Continued) 2.

Verifying with 31 days after removal that laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regula-tory Guide 1.52, Revision 2, March 1978.

3.

Verifying a system flow rate of 39,700 cfm + 10% during.

system operation when tested in accordance with ANSI N510-1975.

b.

After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verify-ing within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52. Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of. Regulatory Guide 1.52, Revisich 2, March 1978.

c.

- At least once.p,er 18 months by verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the system at a flow rate of 39,700 tfm i10%.

d.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEP.A filter banks remove 5 99% of the D0P when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 39,700 cfm i 10%.

e.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove

> 99.95% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 39,700 cfm 110%.

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ARKANSAS - UNIT 2 3/4 9-13

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IREFUELING OPERATIG"5 l

FUEL STORAGE LIl11 TING CONDITION FOR OPERATION j

3.9.12.a Storage in the spent fuel pool shall be restricted to fuel asse b;ies.

having initial enrichment less than or equal to 4.1 w/o U-235. The provisions of Specification 3.0.3 are not applicable.

3.9.12.b Storage in Region 2 (as shown on Figure 3.9.1) of the spent fuel pool shall be further restricted by burnup and enrichment limits specified in Figure 3.9.2.

In the event a checkerboard storage configuration is deemed necessary for a portion of Region 2. vacant spaces adjacent to the faces of j

any fuel assembly which does not meet the Region 2 burnup criteria (Non-Restricted) shall be physically blocked before any such fuel assembly may be placed in Region 2.

This will prevent inadvertent fuel assembly insertion into two adjacent storage locations. The provisions of Specification 3.0.3. are.not j

applicable.

i 1

3. 9.12.c The boron concentration in the spent fuel pool shall be maintained (at all times) at greater than 1600 parts per million.

APPLICABILITY:

During storage of fuel in the ' spent fuel pool.

ACTION:

Y Suspend all actions involving the movement of fuel in the spent fuel pool if'

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it is det&rmined'a"fugl ' assembly has been placed in the incorrect Region until such time as the correct storage location is determined. Move the assembly to l

its correct location before resumption of any other fuel movement.

Suspend all actions involving the movement' of fuel in the spent fuel pool if 1

it is determined the pool boron concentration is less than 1601 ppm, until j

such time as the boron concentration is increased to 1601 ppm or greater.

SURVEILLANCE REQUIREMENTS 4.9.12.a Verify all fuel assemblies to be placed in the spent fuel pool had an initial enrichment of less than or equal to 4.1 w/o U-235 by checking the assemblies design documentation.

4.9.12.b Verify all fuel assemblies to be placed in Region 2 of.the spent fuel pool are within the enrichment and burnup limits of Figure 3.9.2 by checking the assemblies design and burnup documentation.

4. 9.12.c Verify at least once per 31 days the spent fuel pool boron c,oncentra-tion is greater than 1600 ppm.

ARKAiSAS - UNIT 2 3/4 9-14 Amendment No. 43 l

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  • 3/4 9-16 1

Amendment No. 43 l

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i ADMll!!STRATIVE CONTROLS

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6.7 SACETY LIttIT VIOLATION j

6.7.1 The following actions shall be taken in the event a Safety Limit is 1

violated:

i a.

The unit shall be placed in at least HOT STANDBY within one hour.

b.

The Safety Limit violation shall be reported to the Commissio., the i

Vice President, Nuclear Operations and to the SRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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c.

A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the PSC. This report shall describe (1) applicable 4

circumstances preceding the violation (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Commis-sion, the SRC and the Vice-President, Nuclear Operations within 14 l

days of the violation.

6.8 PROCEDURES 4

6.8.1 Written procedures.shall be established, implemented and maintained j

covering the activities referenced below:

a.

The applicable procedufes recommended in Appendix "A" of Regulatory Guide 1.33 Revision.2 February 1978.

b.

Refueling operations.

c.

Surveillance and test activities of safety related equipment.

d.

Security Plan implementation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

g.

Modification of Core Protection Calculator (CPC) Addressable Constants NOTE:

Modification to the CPC addressable constants based on information obtained through the Plant Computer - CPC data link shall not be made without prior approval of-the Plant Safety Committee.

h.

New and spent fuel storage.

6.8.2 Each procedure of 6.8.1 above, and changes thereto, shall be reviewed by the PSC and approved by the General Manager prior to implementation and reviewed periodically as set forth in administrative procedures.

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ARKANSAS - UNIT 2,

6-13 Amendment No.: 5. 77, 24. 25, 43

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ADMINISTRATIVE CONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made pro-vided:

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a.

The intent of the original procedure is not altered.

b.

The change is approved by two members of the plant management l

staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c.

The change is documented, reviewed by the PSC and approved by the General Manager within 14 day of implementation.

l 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.

1 STARTUP REPORT h

6.9.1.1 A summary report of plant startup and power escalation testin shall be submitted following (1). receipt of an operating license, (2) g amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption

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or conmencement of commercial'. power operation, or (3) 9 months following initial criticality, whichever is earliest.

If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or conmencement of commercial power operation), supplementary reports shall b'e submitted at least.

every three months until all three events have been completed.

ARKANSAS - UNIT 2 6-14 Amendment No. 5

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I 3/4.9 REFUELING OPERATIONS

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BASES i

3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent with the initial conditions assumed for the boron dilution incident in the accident analyses.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core; 3/4.9.3 D'ECAY TIME The minimum. requirement for reactor subcriticality prior to movement of irradiated fuel assenblies in the reactor pressure vessel ensures that suffi-cient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

The minimum requirement for reactor subcri.ticality prior to i..ovement of more than 70 irradiated fuel assemblies to the spent fuel pool ensures that I

sufficient time has elapsed to allow radioactive decay of the short lived fission products such that the heat generated will not exceed the cooling capacity of the spent fuel pool cooling system. This decay time and total assembly limitation is conservatively within the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT PENETRATIONS l

The requirements on containment penetration closure and OPERABILITY of the containment purge and exhaust system HEPA filters and charcoal adsorbers ensure that a release of radioactive material within containment will be restricted from leakage to the environment or filtered through the HEPA filters and charcoal adsorbers prior to discharge to the atmosphere. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of contain.

ment pressurization potential while in the REFUELING MODE. Operation of the contaimnent purge and exhaust system HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

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ARKANSAS - UNIT 2 B 3/4 9-1 Amendment No. 43

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REFUELING OPERATIONS BASES 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 REFUELING MA. CHINE OPERABILITY The OPERABILITY requirements for the refueling machine ensure that:

1) the refueling machine will be used for movement of CEAs with fuel assemblies and that it has sufficient load capacity to lift a fuel assembly, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction o'n movement of loads in excess of the nominal weight of a fuel assembly, CEA and associated handling tool over other fuel assemblies in the storage pool ensures 'that in the event this. load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

This assumption is consistent with the activity release assumed in the accident analyses.

J/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION l

The requirement that at least one shutdown cooling loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a baron dilution incident and prevent boron stratification.

The requirement to have two shutdown cooling loops OPERABLE when there is operating shutdown cooling loop will, core ensures that a single failure of the less than 23 feet of water above the not result in a complete 'soss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling, ~thus in the event of a failure of the operat'ng shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.

ARKANSAS - UNIT 2 B 3/4 9-2 AmendmentNo.h,29 w

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REFUELING OPERATIONS BASES 3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM The limitations on the fuel handling area ventilation system ensure that all radioactive materials released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorbers prior to. discharge l

to the atmosphere. The operation of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

3/4.9.12 FUEL STORAGE Region 1 of the spent fuel storage racks is designed to assure fuel assemblies of less ~than'of equal to 4.1 w/o U-235 enrichment will be main-tained in a subcritical array with K,ff 1 95 in unborated water. These 0

conditions have been verified by* criticality analyses.

Region 2 of the spent fuel storage racks is designed to assure fuel assemblies within the burnup and initial. enrichment limits of Figure 3.9.2

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will be maintained in a suberitical array with K,ff 10.95 in unborated water.

These conditions have been verified by criticality analyses.

The requirement for 1600 ppm boron concentration is to assure the fuel assemblies will be maintained in a subcri.tical array with K,ff 1 0.95 in the event of a postulated accident.

I ARKANSAS - UNIT 2 B 3/4 9-3 knendment No. 43 l,

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