ML19352B114

Amend 56 to License DPR-51,modifying App a Tech Specs Dealing W/Reactor Decay Heat Removal Capability

ML19352B114
Person / Time
Site:	Arkansas Nuclear
Issue date:	05/20/1981
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Arkansas Power & Light Co
Shared Package
ML19352B115	List: ML19352B114 ML19352B117 ML19352B120
References
DPR-51-A-056 NUDOCS 8106030126
Download: ML19352B114 (12)

Text

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UNITED STATES y~

t NUCLEAR REGULATORY COMMisslON WASHINGTON, D. C. 20666

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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.56 License No. DPR-51 1.

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

A.

The application for amendment by Arkansas Power and Light Company (the licensee) dated October 31, 1980, as supplemented January 30, 1981, complies with the standards and requirements 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 Comissi,on; 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 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 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

9I06030/Bd3

y 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. DPR-51 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices ~

'A and B, as revised through Amendment No. 56, are hereby inqorporated 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 6J.

Johh: F. Stolf, Chie -

Op rating Reactors Br ch #4 vision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 20,1981 e

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ATTACtENT TO LICENSE AMEN 0 MENT NO. 56 FACIL'TY OPERATING LICENSE NO. OPR-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 emendment number and contain vertical lines indicating the area of change.

Pages 11 16 16a (new page) 17 58 59 59a 59b (new page) 110aa (new page)

SECTION TITLE PAGE 4.

SURVEILLANCE STANDARDS 67 4.1 OPERATIONAL SAFETY ITEMS 67 I

4.2 REACTOR COOLANT SYSTEM SURVEILLANCE 76 4.3 TESTING FOLLOWING OPENING 0F SYSTEM 78

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l 4.4 REACTOR BUILD!.NG 79 l

4.4.1 Reactor Building Leakage Tests 79 4.4.2 Structural Interrity 85 4.5 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 92 l

4.5.1 E:nergency Core Cooling Systems 92 4.5.2 Reactor Building Cooling Systems 95 4.6 AUXILIARY ELECTRICAL SYSTEM TESTS 100 4.7 REACTOR CONTROL ROD SYSTEM TESTS 102 4.7.1 Control Rod Drive System Functional Tests 102 4.7.2 Control Rod Program Verification 104 4.8 EMERGENCY FEEDWATER PUMP TESTING 105.

4.9 REACTIVITY ANOMALIES 106 4.10 CONTROL ROOM EMERGENCY AIR CONDITIONING AND ISOLATION SYSTEM SURVEILLANCE

,107 4.11 PENETRATION ROOM VENTILATION SYSTEM SURVEILLANCE 109 4.12 HYDROGEN PURGE SYSTEM SURVEILLANCE 109b 4.13 EMERGENCY COOLING POND 110a 4.14 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 110b 4.15 AUGMENTED INSERVICE INSPECTION PROGRAM FOR HIGH ENERGY LINES OUTSIDE OF CONTAINMENT 110e 4.16 SHOCK SUPPRESSORS'(SNUBBERS) 110e 4.16.1 Hydraulic Shock Suppressors 110e 4.17 FUEL HANDLING AREA VENTILATION SYSTEM SURVT.ILLANCE 110h 4.18 STEAM GENERATOR TUBING SURVEILLANCE 1101 4.19 FIRE _ DETECTION INSTRUMENTATION 110p 4.20 FIRE SUPPRESSION WATER SYSTEM 110q 4.21 SPRINKLER SYSTEMS 110c 4.22 CONTROL ROOM AND AUXILIARY CONTROL ROOM BALON SYSTEMS 110u 4.23 FIRE HOSE STATIONS 110v 4.24 PENETRATION FIRE BARRIERS 110w 4.25 REACTOR BUILDING PURGE FILTRATION SYSTEM 110x 4.26

tEACTOR BUILDING PURGE VALVES 110z_

4.27 DECAY HEAT REMOVAL 110aa l

S.

DESIGN FEATURES 111 5.1 SITE 111 5.2 REACTOR BUILDING 112 5.3 REACTOR 114 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 116 6.

ADMINISTRATIVE CONTROL 3 117 6.1 RESPONSIBILITY 117 6.2 ORGANIZATION 117 6.3 FACILITY STAFF QUALIFICATIONS 117 6.4 TRAINING 117 6.5 REVIEW AND AUDIT 117 i

6.6 REPORTABLE OCCURRENCE ACTION 126 6.7 SAFETY LIMIT VIOLATION 126 6.8 PROCEDURES 127 s

6.9 RECORD RETENTION 128 6.10 RADIATION PROTECTION PROGRAM 129 6.11 HIGH RADIATION AREA 129 6.12 REPORTING REQUIREMENTS 140 6.13 ENVIRONMENTAL QUALIFICATION 147 24 30. 34, 11 2,44,,EE, 56 Amend =ent No.

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3.

LIMITING CONDITIONS FOR OPERATION 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the reactor coolant system.

Objective To specify those limiting conditions for operatic: of the reactor coolant system which must be met to ensure safe reactor cperations.

3.1.1 Operational Components Specification

.3.1.1.1 Reactor Coolant Pumps A

Pump combinations permissible fo'r given power levels shall be as shown in Table 2.3-1.

B.

The boron concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay Leat removal pump is circulating reactor coolant.

3.1.1.2 Steam Generator A.

Two steam generators shall be operable whenever the

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reactor coolant average temperature is above 280*F.

3.1.1.3 Pressurizer Safety Valves A.

The reactor shall not remain critical unless both pres-surizer code safety valves are operable.

B.

When the reactor is subcritical, at least one pres-surizer code safety valve thall be operable if all re-actor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Boiler and Pressure Vessel Code,Section III.

3.1.1.4 Reactor Internals Vent Valves The structural integrity and operability of the reactor in-ternals vent valves shall be maintained at a level consistent with the acceptance criteria in Specification 4.1.

3.1.1.5 Reactor Coolant Loops A.

With the reactor coolant average temperature above 280'F, the reactor coolant loops listed below shall be operable:

Amendment No. 21,56 16 s

o.

1.

Reactor Coolant Loop (A) and at least one associated reactor coolant pump.

2.

Reactor Coolant Loop (B) and at least one associated reactor coolant pump.

Otherwise, restore the required loops to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or reduced the reactor coolant average temperature to less than or equal to 280*F within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B.

With the reactor coolant average temperature above 280*F, at least one of the reactor coolant loops listed above shall be in operation.

Otherwise, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and is-mediately initiate corrective action to return the required loop to operation.

3.1.1.6 Decay Heat Removal With the reactor coolant average temperature at or below 280*F, but the reactor above the refueling shutdown condition, at least two of the coolant loops listed below shall be operable, and at least one loop shall be in operation:*

1.

Reactor Coolant Loop (A) and its associated steam generator and at least one associated reactor coolant pump.

2.

Reactor Coolcat Loop (B) and its associated steam generator and at least one associated reactor coolant pump.

3.

Decay Heat Removal Loop (A)**

4.

Decay Heat Removal Loop (B)**

A.

With less'ttan the above required coolant loops OPERABLE,-

immediately initiate corrective action to return the required coolant loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

B.

With no coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

• All reactor coolant pumps and decay heat removal pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron con-centration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

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• • The normal or emergency power source may be inoperable when the reactor is in a cold shutdown condition.

Anendment No. 56 16a

BASES:

The plant is designed to operate with both reactor coolant loops and at least one reactor coolant pump per loop in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients. (1)

Whenever the reactor coolant average temperature is above 280*F, single failure considerations require that two loops be operable.

The decay heat removal system suction piping is designed for 300'F, thus, the system can remove decay heat when the reactor coolant system is below this temperature. (2,3)

One pressurizer code safety valve is capable of preventing overpres-surization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sour whicharepumpenergy,pressurizerheaters,andreactordecayheat.{g3 Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabili-ties.

The {gje safety valves prevent overpressure for a rod withdrawal accident.

The pressurizer code safety valve lift set point shall be set at 2,500 psig i 1 percent allowance for error and each valve shall be capable of relieving 300,000 lb/h of saturated steam at a pressure not greater than 3 percent above the se,t pressure.

The internals vent valves are provided to relieve the pressure generated by steaming in the cere following a LOCA so that the core remains sufficiently covered.

Inspection and manual actuation of the internal vent valves (1) ensure operability, (2) ensure ~that the valves are not oper. during normal operation, and (3) demonstrate that the valras begin to open and are fully open at the forces equivalent to the differential pressures assumed in the safety analysis.

REFERENCES (1) FSAR, Tables 9-10 and 4-3 through 4-7 (2) FSAR, Section 4.2.5.1 and 9.5.2.3 (3) FSAR, Section 4.2.5.4 (4) FSAR, Section 4.3.10.4 and 4.2.4 (5) FSAR, Section 4.3.7 Amendment No. 21, 56 17

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38 TJEL 10ADIEG AHD REFELING Applicability Applies to fuel loading and refueling operations.

t Objective To assure that fuel loading, refueling and fuel handling operations are per-formed in a responsible manner.

Specificat. ion 3 8.1 Radiation levels in the reactor building refueling area shall be mon-itored by instrument RE-8017 Radiation levels in the spent fuel storage area shall be monitored by instrument RE-8009 If any of these instru-ments become inoperable, portable survey instru=entation", havin'g the appropriate ranges and, sensitivity to fully protect individuals involved in refueling Operation, shall be used until the per=anent instru=entation is. returned to service.

3 8.2 core suberitical neutron flux shall be continuously monitored by at least two neutron flux monitors, each with continuous indication avail-able, whenever core geometry is being changed. When core geometry is not deing changed, at least one neutron flux monitor shall be in service

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3.8.3.a.

At least one decay heat remova] loop shall be in operation.*

Otherwise, suspend all operations involving an increase in the reactor 7

decay heat load or a reduction in boron concentration of the reactor coolant system, and close all containment penetrations providing access from the contaDurf atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

When the water level above the top of the irradiated fuel assemblies seated within the reactor pressure vessel is less than 23 feet, two decay heat removal loops shall be operable.**

Otherwise, immediately initiate corrective action to return the required loops to operable status as soon as possible.

3 8.4 During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concent-ation shall be maintained at not less than that required for refueling shutdown.

I 3.8 5 Direct co==unications between -

control room and the refueling per-l sonnel in the reactor building =aal.1 exist whenever changes in core geometry are taking place.

• The decay heat removal loop may be removed from operation for up to t

,f I hour per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period during the performance of core alterations.

• • The normal or emergency power source may be inoperable for each shut-down cooling loop.

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l Amendment No. 56 58 L

o 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 batch cover shall be in place with a minimum of four bolts securing the cover to the sealing surfaces.

387 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 i

the bridges within the fuel transfer canal, a minimum of 10 feet sep-aration shall be maintained between the assemblies at all times.

389 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.

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.

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

3.8.12 All fuel handling in the Auxiliary Building shall cease upon notifi-cation of the issuance of a tornado watch for Pope, Yell, Johnson, or Logan countiesi in Arkansas. Fuel handling operations in progress vill be completed to the extent necessary to place the fuel ha,ndling bridge and crane in their normal' parked and locked poJition.

3.8.13 N 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 acecmpanied by a vind velocity of 2 =/sec. In addi-tion, the railroad spur door of the Turbine Building shall be closed and the fuel handling a ea ventilation system shall be in operation.

3.8.14 For the maximum fuel pool heat lo'ad capacity (i.e., seven reload batches (413 assemblies) stored in the pool at the time of discharge l

I of the full core) the full core to be discharged shall be cooled in i

che reactor vessel a mini =um of 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> prior to discha::ge.

~~3.8.15 Loads'in' excess of 1000' pouads shall be prohibited from travel over fuel assemblies in the storage pool.

3.8.16 The, spent fuel shipping cask shall not be carried by the j

auxiliary building crane pending the evaluation of the spent fuel cask drcp accident and the crane design by AP&L an( NRC review and approval.

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l Amendment No. 16, 17, M,56 59 i

J 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 described in Section 9.7 of the FSAR i

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 r

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 opera-tion 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 maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.I3) l The requirement to have two decay heat removal loops operable when there is 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 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 operat-ing decay heat removal loop, adequate time is provided to initiate emergency procedures to cool the core.

The shutdown margin indicated in Specification 3.8.4 will keep core j

subcritical, even with all control rods withdrawn from the core The boron concentration will be maintained above 1,800 ppe. Although this 4

concentration is sufficient to maintain the core k 0.99 if all the controlrodswereremovedfromthecore,onlyafewbo<trolrodswillbe f

n removed at any one time during fuel shuffling and replacement. The k with all rods in the core and with refueling boron concentration i,gg s

approximately 0.9.

Specification 3.8.5 allows the control room operator to inform the reactor building personnel of any impending unsafe condi-tion detected from the main control board indicators during fuel move-ment.

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 signifi-cant fisrion 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 the safety analysis for the fuel handlingaccidentwasggjedontheassumptionthatthereactorhadbeen shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Specification 3.8.14, which requires cooling of the full core for 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> prior to discharge to the spect fuel pool when seven reload b stehes are already stored in the pool, is necessary to assure that the m ximum design heat 1 cad of the spent fuel pool cooling system will not be exceeded.

1 393 Amendment No.17, 56

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O Specification 3.8.15 will assure that damage to fuel in the spent fuel pool will not be caused by dropping heavy objects onto the fuel. Ad-ministrative 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 3.8.16 is completed.

Specification 3.8.16 assures that the spent fuel caskairop accident cannot 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 3.8.16 shall be deleted.

REFEREi4CES (1) FSAR, Section 9.5 (2) FSAR, Section 14.2.2.3 (3) F3AR, Section 14.2.2.3.3

'l Amendment No. 56 59b

4.27 DECAY llEAT REMOVAL APPLICABILITY Applies to surveillance of the decay heat removal system and to the reactor coolant loops and associated reactor coolant pumps as needed for decay heat removal.

OBJECTIVE To assure the operability of the decay heat removal system and the reactor coolant loops as needed for decay heat removal.

SPECIFICATION 4.27.1 The required reactor coolant pumps shall be determined operable once per seven (7) days by verifying correct breaker aligsments and indicated power availability.

4.27 2 The required decay heat removal loop (s) shall be determined operable per Specification 4.2.2.

4.2;.3 The required steam generator (s) shall be determined operable by verifying the secondary side water level to be ?. 20 inches on the startup range at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.27.4 The required reacter coolant loop (s) shall be ar.t-"'i'2d operable by verifying the required loop (s) to 'ue >u opecation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.27.5 The required decay heat removal loop shall be determined to be in operation at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Amendment No. 56 110aa

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