Text

Proposed Changes to Tech Specs Sections 3,4,5 & 6 Re Spent Fuel Pool Expansion
	ML20079L665
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/17/1983
From:	; ARKANSAS POWER & LIGHT CO.
To:	;
Shared Package
ML20079L652	List: 0CAN028302, Application for Amend to Licenses DPR-51 & NPF-6,changing Tech Specs Sections 3,4,5 & 6 Re Spent Fuel Pool Expansion. Class IV Fee Encl; ML20079L665;
References
	NUDOCS 8302230408
	Download: ML20079L665 (17)
	v • d • e

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3.8.6 During the handling of irradiated fuel in the reactor building 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.

3.8.7

. Isolation valves in lines containing automatic containment isolation valves shall be operable,-or at least one shall be closed.

3.8.8 When two irradiated fuel assemblies are being moved simultaneously 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 shall cease; action shall 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 reactor building purge isolation system, including the radiation monitors shall be tested and verified to be operable within 7 days prior to refueling operations.

The provisions of Specifications 3.0.3 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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

Tse provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

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

Fuel handling operations i

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 Auxiliary 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 3.0.3 and 3.0.4 are not applicable.

3.8.14 Loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the storage pool.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

Amendment No. 70. 77, 36, 57 59 8302230400 830217 PNtADOCK05000g

e..

t 3.8.15 The spent fuel shipping cask:shall-not' be carried by the 1 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 shall be maintained (at all times) at greater than 1600 parts' per million.

I'

~ BASES

{

Detailed written procedures will be.avai.lable for use by refueling i

personnel.

These procedures, the above specifications, and the design of i-the fuel handling equipment as described 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 indicat. ion of an unsafe condition.'

The requirement that at least one decay heat removal loop be in operation ensures that (1) sufficient coolinr capacity is available to remove decay-4 heat and maintain the. water in the reactor: pressure vessel at the refueling ten.perature (normally 140*F), and (2) ' sufficient coolant circulation is maintainedthroughthereactorcoretominimizethegfectsofaboron F

dilution incident and prevent boron stratification.

l The requirement to have two decay heat removal loops operable when there is L

less than 23 feet of water above the core, ensures that a single failure of the operating decay heat removal loop will not result in a comolete 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 temoval F

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

TheshutdownmarginindicatedinSpecification3.8.4willkeeptgcore subcritical, cven with all control rods withdrawn from the core j

Although the refueling boron concentration ~is sufficient to maintain the core k 6 0.99 if all the control rods were removed from the core, only a l

fewcobolrodswillberemovedatanyonetimeduringfuelshufflingand i

Amendment No. 17, 56, 57 59a

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.replacemenL ^The k with.all rods in the core and with refueling boron-concentrationisaphximately0.9.

Specification'3.8.5 allows the control

~

room operator to inform the reactor building personnel of any impending.

unsafe condition detected from the main control. board indicators during fuel

' movement.

The specification 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 g ed'on the assumption that the reactor.had been shutdown for.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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 'amage to fuel.in the spent.. fuel pool d

will not be 1 sed by dropping heavy objects onto the fuel.

Administrative controls'w'..: 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 del _eted.

l

~

Specifications 3.8.16 and 3.8.17 assure fuel enrichment and fuel burnup j

limits assumed in-the spent fuel safety analyses will not be exceeded.

j Specification 3.8.18 assures the boron concentration in the spent fuel pool

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

l REFERENCES (1)

FSAR, Section 9.5 (2)

FSAR, Section 14.2.2.3 (3)

FSAR, Section 14.2.2.3.3 Amendment No. 56, 57 59b

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0.0M. 0 2.0 3.0 41 Initial Assembly Average Enrichment (w/o U-2 5) 59d Amendment No.

'T 5.4 NEW AND SPENT FUEL STORAGE FACILITIES

' Applicability Applies 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 occur.

Specification 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 lessthan.9eveniffloodedwithunboratedwater, base 8$n fuel with an enrichment of 3.5 weight percent U235.

2.

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

~5.4.2 Spent 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 i.

r.

1 e

Amendment No. 1Gf 116 i

. a.

The. facility shall be placed_in at-least hot shutdown within one hour.

b.

The Nuclear Regulatory Commission shall be notified and a-report. submitted pursuant to the requirements of 10 CFR 50.36 and Specification 6.12.3.1.

4

- 6. 8 -

PROCEDURES

- 6.8.1 Written' procedures shall be established, imp emented.and maintained' covering the activities references below:

a.

The applicable procedures recommended.in_Appen'ix "A" of d

Regulatory Guide 1.33, November, 1972.

I 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.

New and' spent fuel storage locations.

l.

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 1

administrative procedures.

i 6.8.3 Temporary changes to procedures of 6.8.1 above may be'made j

provided:

a.

The intent of the original procedure is not altered.

i-b.

The change is approved by two members of the plant 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 days of implementation.

i b

i 4

l

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Amendment No. 16, 30, 34, 37

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distrubution - Operating.......................

3/4 8-6 A.C. Distribution - Shutdown........................

3/4 8-7 D.C. Distribution - Operating.......................

3/4 8-8 D.C. Distribution - Shutdown........................

3/4 8-10 Containment Penetration Conductor Overcurrent Protection Devices................................

3/4 8-11 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION.................................

3/4 9 3/4.9.2 INSTRUMENTATION.....................................

3/4 9-2 3/4.9.3 DECAY TIME AND SPENT FUEL STORAGE...................

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

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

3/4 9-6 3/4.9.6 REFUELING MACHINE 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 P0OL WATER LEVEL.........................

3/4 9-11 3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEM...............

3/4 9-12 3/4.9.12 FUEL STORAGE........................................

3/4 9-14 ARKANSAS - UNIT 2 VIII Amendment No.

INDEX LIMITIH9 CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN.....................................

3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUT_DN LIMITS...............,..............................

3/4 10-2 3/4.10.3 REACTOR COO LANT 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 i

K ARKANSAS - UNIT 2 VIIIa Amendment No.

REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3.a The reactor shall be subcritical for at least~ 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months prior to discharge of more than 70 assemblies to the spent fuel pool.

APPLICA3ILITY:

During movement of irradiated fuel in the reactor-pressure vessel.

ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.

With the reactor subcritical for less than 175 hours0.00203 days 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months , suspend all operations involving movement of more than 70 fuel assemblies from the reactor pressure 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 subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verification of the date and time of subcriticality prior to movement 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 0.0486 hours 2.893519e-4 weeks 6.65875e-5 months by verification of the date and time of subcriticality prior to movement of the 71st irradiated fuel assembly from the reactor pressure vessel to the spent fuel pool.

ARKANSAS - UNIT 2 3/4 9-3 Amendment No.

~

REFU LING OPERATIONS-FUEL" STORAGE.

LIMITING CONDITION FOR OPERATION 3.9.12.a 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 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 j

pool shall be further restricted by burnup and enrichment limits specified in Figure 3.9.2.

The provisions of Specification 3.0.3 are not applicable.

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:

Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined a _ fuel assembly has been placed in the incorrect Region until such time as the' correct storage location is determined.

Move the i

i assembly to its correct location before resumption of any other. fuel movement.

Suspend all actions. involving the movement of fuel in the spent fuel pool if l

it is determined the pool boron' concentration is less than 1601 ppm, until such time as the boron concentration is increased to 1601 ppm or greater.

i 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.

i i-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 assembly's-design and burnup documentation.

i 4.9.12.c Verify at least once per 31 days the spent fuel pool boron concentration is greater than 1600 ppm.

I a

ARKANSAS - UNIT 2 3/4 9-14 Amendment No.

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3/4 9-15 Amendment No.

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rs 0.041.0 2.0 3.0 4.1 Initial Asse.mbly Average Enrichment (w/o U-235 )

ARKANSAS-Unit 2 3/4 9-16 Amendment No.

ADMINISTRATIVE CONTROLS 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Liinit is violated:

a.

The unit shall be placed in at least HOT STANDBY within o 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

b.

The Safety Limit violation shall be reported to the Commission, the Vice-President, Nuclear Operations and to the SRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~

c.

A Safety Limit Violation Report shall be prepared.

The report shall be reviewed by the PSC.

This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation _upon facility comporents, systems or structures, and (3) corrective action taken to prevent recurrence.

d.

The Safety Limit Violation Report shall be submitted to the Commission, the SRC and the Vice-President, Nuclear Operations l

within 14 days of the violation.

6.8 PROCEDURES 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

.The applicable procedures recommended in Appendit "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 implerrentation.

e.

Emergency Plan implementation.

f.

Fire Protection Program implementation.

g.

Modification of Core Portection 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 locations.

l ARKANSAS - UNIT 2 6-13 Amendment No. 5, 17, 24 25

=

~

6.8.2 Each proce' dure of 6.8.1 above, and changes thereto, shall be reviewed by the PSC and approved by the General Manager prior to imp!nmentation and reviewed periodically as set forth in adr:inistratvie procedures.

b j-l ARKANSAS - UNIT 2 6-13a Amendment'No.

4 REFUELING OPERATIONS 3/4.9 REFUELING OPERATIONS RAtF9 RAiFs 3/4.9.9 and 3/4.9.10 WATER LEVEL-REACTOR VESSEL AND SPENT FUEL POOL WATER 3/4.9.1 BORON CONCENTRATION LEVEL The limitations on reactivity conditions during REFUELING ensure that:

The restrictions on minimum water level ensure that sufficient water

1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a depth is available to remove 995 of the assumed 10% iodine gap activity uniform boron concentration is maintained for reactivity control in the released from the rupture of an irradiateu fuel assembly. The minimum water water volume having direct access to the reactor vessel. These limitations depth is consistent with the assumptions of *.he accident analysis are consistent with the initial conditions assumed for the boron dilation incident in the accident analyses.

3/4.9.11 FUEL HANDLING AREA VENTILATION SYSTEN 3/4.9.2 INSTRUNENTATION The limitations on the fa.el handling area ventilation system ensure that all re.dioactive materials released from an irradiated fuel assembly The OPERABILITY of the source range neutron flux monitors ensures that will be filtered througn the HEPA filters and charcoal adsorbers prior to redundant monitoring cambility is available to detecs changes in the disenarge to the atmosphere. The operation of this system and the resulting reactivity condition of the core, iodine removal capacity are consistent with the assumptions of the accident analyses.

3/4.9.3 DECAY TIME 3/4.9.12 FUEL STORAGE The minimum requirement for reactor subcriticality prior to movement of fr-adiated fuel assemblies in the reactor pressure vessel ensures that Region 1 of the spent fuel storage racks is designed to assure fuel sufficient time has elapsed to allow the radioactive decay of the short assemblies of less than or equal to 4.1 w/o U-235 enrichment will be lived fission products. Thl6 decay time is consistent with the assumptions maintained in a subcritical array with K 50 These used in the accident analyses.

conditions have been verified by criticaTIfy an.95 in unborated water.

alyses.

The minimum requirement for reactor subcriticality prior to movement of Region 2 of the spent fuel storage racks is designed to assure fuel more than 70 irradiated fuel assemblies to the spent fuel pval ensures that assemblies within the burnup and initial enrichment limits of Figure 3.9 2 sufficient time has elapsed to allow radioactive decay of the short Ilved 4

will be maintained in a subcritical a+ ray with K 50.95 in unborated fission products such that the heat generated will not exceed the cooling 6

water. Theseconditionshavebeenverifiedbycff(icalityanalyses.

capacity of the spent fuel pool cooling system. This decay time and total assembly limitation is conservatively within the assumptions used in the The requirement for 1600 ppe boron concentration is to assure the fuel accNent analyses.

assemblies will be saintained in a subcritical array with K,gy 50.95 in the event of a postulated accident.

3/4.9.4 CONTAINNENT PENETRATIONS The requirements on containment penetration closure and OPERA 8ILITY of the containment purge and exhaust system HEPA filters and charcoal adsorbers i

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 prier 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 i

containment pressurization potential while in the REFUELING MODE. Operation L

3 1

of the containment purge and exhaust systaa HEPA filters and charcoal adsorbers and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

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

ARKANSAS - UNIT 2 8 3/4 9-1 Amendment No.

i n

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