ML19326C354

Amend 4 to License DPR-51,changing Tech Specs Re Main Steam Pressure & Channel Operation for Steam Line Break Instrumentation & Control Sys

ML19326C354
Person / Time
Site:	Arkansas Nuclear
Issue date:	09/11/1975
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML19326C351	List: ML19326C350 ML19326C354 ML19326C356 ML19326C359
References
NUDOCS 8004220899
Download: ML19326C354 (13)

Text

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ARKANSAS POWER AND LIGrr COMPANY bb DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE - UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE

[f Amendmea: No. 4' License No. DPR-51

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1.

The Nuclear Regulatory Connission (the Coranission) has found that:

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A.

%e application for amendment by Arkansas Power and Light Company (the licensee) dated April 17, 1975, complies with the standards and requirements of the Atomic Energy Act of i

1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

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B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and rpsilations of the Commission; t'~

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; and D.

The issuance of this amendcent will not be inimical to the cori: mon defense and security or to the health and safety of the public.

2.

Accordingly, the license is amended by a chango to the Technical Specifications as indicated in the attachment to this license a::,endr,ent and Paragraph 2.c(2) of Facility License No. DPR-51 is hereby amendod to read as follows:

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"(2). Technical Specifications

=.5 The Technical Specifications contained in Appendices

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A and B, as revised, are hereby incorporated in the au license. The licensee shall operate the facility in accordance with the Technical Specifications, as revised EE by issued changes thereto through Change No. 4."

3.

This license amendment is effective as of the date of its issuahce.

FOR'niE NUCLEAR REGULATORY COFNISSION Or! inal Sicned h7 ?

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Denni3 L. Ziemann

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Dennis L. Ziemann, Chief Operating Reactors Branch #2 E~

Division of Reactor Licensing

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Attachment:

Change No. 4 to the Technical Specifications E

Date of Issuance:

8Ep 111975 9..

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ATTAQi!ENT TO LICENSE AMENDMENT NO. 4 I!!!

O!ANGE NO. 4 TO Tile TEGINICAL SPECIFICATIONS i..gg FACILITY OPERATING LICENSE NO. DPR-S1 FEE DOCKET NO. 50-313 Delete pages 39a, 40, 41, 42, 43a, 44, 45b, 46, 72 and 73a from the.

Appendix A Tedmical Specifications and insert the attached replaces:aent

. :.:.:s pages. The changed areas on the revised pages are shown by a marginal line.

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- The penetration room ventilation system consists of two independent, full capacity, 100% redundant trains.

the other train must be operable. g one train is removed from operation, REFERENCES (1) FSAR, Section 14.2 5 (2) FSAR, Sectica 3.2 (3) FSAR, Secozon 9 5.2 -

(h-) FSAR, Section 9.3.1 (5) FSAR, Section 6.5 9

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3.4 STEAM AND POWER CONVERSION SYS'IUf^

Applicability Applies to the turbine cycle components fbr removal of reactor decay heat.

Objective To specify minimum conditions of the turbine cycle equipment necessary to assure the capability to remove decay heat from the reactor core.

Specifications 3.4.1 The' reactor shall not be heated, above 2800F unless the.fo' llowing conditions are met:

1.

Capability to remove a decay heat load of 5% full reactor power by at least one of the following means:

A condensate pump and a main feedwater pump, using turbine

a..

by-pass valve, b.

A condensate pump and the auxiliary feedwater pump using turbine by-pass valve.

2 Fourteen of the steam system safety valves are operable.

3.

A minimum of 16.3 ft. (107,000 gallons) of water is avail-abic in the condensate storage tank.

4.

Both emergency feedwater pumps are operable.

5.

Both main steam block ' valves and both main feedwater isolation valves are operable.

6.

The emergency feedwater valves associated with Specification 3.4.1.4 shall be operable.

3.4.2 The Steam Line Break Instrumentation and Control System (SLBIC) shall be operable when main steam pressure exceeds 700 psig and 4

shall be set to actuate at 600 25 psig-.

3.4.3 Components required by Specification 3.4.1 and 3.4.2 to be operable shall not be removed from service for more than 24 consecutive hours.

If the system is not restored to meet the requirements of Specification 3.4.1 and 3.4.2 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the reactor shall be placed in the hot 14 shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

If the requirements of Specifi-

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3.4.1 and 3.4.2 are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor

[4 shall be placed in the cold shutdown condition 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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Bases The feedwater flow required to remove decay heat corresponding to 5% full power with saturated steam at 1065 psia (lowest ' setting of steam safety valve)~ as a function of feedwater temperatu.re is:

Feedwater Temperature Flow

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60 758 90 777-120 799 140

.814 The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280 F.. Feedwater makeup is supplied by operation of a condensate pump and either a main or the auxiliary

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feedwater pump.

In the incredible event of loss of all AC power, feedwater is supplied by the turbine driven emergency fendwater pump which, takes suction from the condensate storage tank.

Decay heat is removed from a steam generator by steam relief through the atmospheric dump valves or safe'ty valves.

Fourteen of the steam system safety valves will relieve the necessary amount of steam for rated reactor power.

The minimum amount of water in the condensate storage tank would be adequate for abcut 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of. operation.

This is based on the estimate of the average emergency flow to a steam generator being 390 gpm.

This operation time with the volume of water specified would not be reached, since the decay heat removal system would be brought into operation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less.

If the turbine driven emergency feedwater pump has not been verified to be operable within 3 months prior to heatup its operability will be verified upon r'eaching hot shutdown conditions.

The SLBIC System is designed to isolate the steam generators to as'sure that only one steam generator will experience uncontrolled blowdown following a

s. team line break.

Normal steam line operating pressures are approximately 900 psig at all power levels, thus operability above 700 psig with actuation 4

at 600 25 psig are appropriate.

The setpoint is based on severe transients in the main steam lines resulting in rapid pressure decays.

, References FSAR, Section 10 41

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t 35 I!!STRUMENTATICIT SYSTEMS 3 5.1 operational Safety Instrumentation Applicability Applies to unit instrumentation and control systems.

• Objectives To delineate the conditions of the unit instru=entation and safety circuits necessary to assure reactor safety.

Specifications 3 5 1.1 Startup and operation are not permitted unless the requirements of Table 3.5.'1-1, columns 3 and b are met.

3.51.2. In the event the nu=ber of protection channels operable falls below the limit given under Table 3.5.1-1, columns 3 and 4, cperation shall be limited as specified in Column 5 3.5.1.3 For on-line testing or in the event of a protection instrument or channel failure, a key operated channel bypass switch associated with each re actor protection channel vill be used to lock the channel trip relay in the untripped state as indicated by a li6 t.

h Only one channel shall be locked in this untripped state at any one time.

Only one channel bypass key shall be accessible for use in the control roon.

3.5.1.h The key operated shutdovn bypass switch associated with each reactor protection channel shall not be used during reactor pcuer operation.

3515 During startup when the intermediate range instruments come on scale, the overlap between the intermediate range and the source range instru-mentation shall not be less than one decade.

If the overlap is less than one decade, the flux level shall be maintained in the source range until the one decade overlap is achieved.

3.5 1.6 In the event that one of the trip devices in either of the sources supplying power. to the control rod drive mechanists fails in the

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untripped state, the power supplied to the rod drive mechanisms

,through the failed trip device shcl1 be manually removed within 30 minutes following detection.

The condition will be corrected and the remaining trip devices shall be tested within eight hours following detection.

If the condition is not corrected and the remaining trip devices are not tested within the eight-hour period, the reactor shall be placed in the hot shutdown condition within an additional four hours.

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m for protective action from a digital ESAS subsystem will not cause that sub-system to t. rip.

The fact that a module has been removed will be continuously annunciated to the operator.

The redundant digital subsystem is still suffi-cient to indicate complete ESAS action.

'Ihe testing schemes of both the RPS and the ESAS enable complete system test-ing while the reactor is operating.

Each channel is capable of being tested independently so that operation of individual channels may be evaluated.

The Automatic Closure and Isolation System (ACI) is designed to close the Decay Heat Removal System (DHRS) return line isolation valves when.the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the DHRS design pressure or when core flooding system isolation valves are opened.

The ACI is designed to permit manual operation of the DHRS return line isolation

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valves when permissive conditions exist.

In addition, the ACI is designed to disallow manuel operation of the valves when permissive conditions do not exist.

Power is normaily supplied to the control rod drive mechanisms from two sepa-rate parallel 480 volt sources.

Redundant trip devices are employed in each of these sources.

If any one of these trip devices fails in the untripped state on-line repairs to the failed device, when practical will be made, and the remaining trip devices will be tested.

Four hours is ample time to test the remaining trip devices and in many cases make on-line repairs.

The Steam Line Break I

.ation and Control System (SLBIC) is designed to automatically closs a Main Steam Block valves and the Main Feedwater Isolation valves upon loss of pressure in either of the 'two main steam lines. [4 The SLBIC is also designed to be reset from its trip pos.ition only when the system is shut down or the Main Steam line pressure is below 650 psig.

l4 REFERENCE FSAR, Section 7.1 A

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Table 3.5.1-1 Instrumentation Limiting conditions for operation (No*.e 6)

REACTOR PROTECTION SYSTEM 1

2 3

4 5

i Operator action No or channels Min.

Min, is conditions of No. of for sys-operable degree of column 3 or k Functional Unit channels tem trio channels redundancy cannot be met i

1.-

Manual punhbutton 1

1 1

O Note 1 2.

Power re.nge instrument channel h

2 3 (Note 4) 1 (Note 4)

Note 1 3.

Intermediate range instrument channels 2

Note 7 1

• O' Notes 1, 2 k'.

Source range instrument channels 2

Note 7 1

0 Notes 1, 2. 3 5.

Reactor coolant temperature instrument h

2 2

1 Note 1

• P channels 6.

Pressure-temperature instiument 4

2

'2 1

Note 1 channels 7.

Flux / imbalance / flow instrument 4

2 '

2 1

Note 1 channels

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Reactor coolant pressure a.

High reactor coolant pressure 4'

2 P

1 Note 1 instrument channels b.

Low reactor coolant pressure 2

2 1

Note 1 instrument channels 9.

Power / number of pumps instrument k

2 2

1 Note 1 channels 10.

High reactor building pressure h

2 2

1 Note 1 channels i-

i Ttblo 3.5.1-1 (Centd) 0111ER SAFETY RELATED SYSTEMS 1

2 3

4 5

No. of Operator action channels Nun.

Min.

If conditions of No. of for sys-operable degree of column 3 or 4

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Functional unit channels tem trip channels redundancy cannot be met 2.

Steam line break instrumentation control system (SLBIC) a.

SLBIC Control 6 Logic channels 2

1 2

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Notes:

1.

Initiate a shutdown using norral operating instructions and place the reactor in the hot shut-dmen condition if the requirements of Columns 3 and 4 are not met within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2 When 2 of 4 power range instrument channels are greater than 10% rated power, hot shutdown is not required.

g; 3.

When 1 of 2 intermediate range instrument channels is greater that 10 I amps, hot shutdown is cr not required.

For channel ' esting, calibration, or maintenance, the minimum number of operable channels may 4

t be two and a degree of redundancy of one for a maximum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, after which Note-1 applies.

S.

If the requirements of Columns 3 or 4 cannot be met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, place the reactor in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

j 6.

The minimum number of operabic channels may be reduced to 2, provided that the system is reduced to 1 out of 2 coincidence by tripping the remaining channel.

Otherwise, Speci-fication 3.3 shall apply.

i 7.

These channels initiate control rod withd' awal inhibits not reactor trips at <10% rated power.

r Above 10% rated power these inhibits are bypassed.

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If any one component of a digital subsystem is inoperabic, the entire digital subsystem is considered inoperable.

llence, the essociated safety features are inoperable and Speci-j fication 3.3 applies.

t 9.

The minimum number of operable channels may be reduced to one and the minimum degree of 4

redundancy to zero for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after which Note 1 applies, 4

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,s 352 Control Rod Group and Power Distribution Limits Apolicability This specification applies to power distribution and operation of control rods during power operation.

Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core suberiticality after a reactor trip.

Specification 3 5 2.1 The available shutdown margin shall be not less than 15 Ak/k with the highest worth control rod fully withdrawn.

3.5 2.2 Operation with inoperable rods:

1.

Operation with more than one inoperable rod, as defined in Specification h.7.1 and 4.7 2.3, in the safety or regulating rod groups shall not be permitted.

2.

If a control rod in the regulating or safety rod groups is declared inoperable in the withdrawn position as defined in Specification h.7.1.1 und 4.7.1.3, an evaluation shall be initiated i==ediately to verify the existance of 15 Ak/k available shutdown carcin.

Boration. cay be initiated either to the vorth of the inoperable rod or until the regulating and transient rod groups are withdrawn to the limits of Specification 3.5.2 5.3, whichever occurs first.

Si=ultan-cously a program of exercising the retaining regulating and safety rods shall b( initiated to verify operability.

3 If within one (1) hour of determination of an inoperable rod as defined in Specification h.71, it is not determined that a 1% Ak/k available shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is established, k.

Following the determination of an inoperable rod as defined in Specification h.71, all remaining' rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised vackly until the rod problem is solved.

5 If a control roe in the regulating or safety rod groups is declarca inoperable per k.7 1.2, power shall be reduced to C0% of the thermal power allovable for the reactor cool-I ant pump combination.

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s Table 4.1-1 (Cont'd)

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Channel Description Check Test Calibrate Remarks 30.

Decay !! cat Removal S(l)(2)

M(1)(3)

R' (1) Includes RCS Pressure Anal'og System Isolation Valve Channel Automatic Closure And Interlock System (2) Includes CFT Isolation Valve

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Position 1

(3) Shall Also Be Tested During Refueling Shutdown Prior to Re-pressurization at a_ pressure greater than 300 but less than 420 psig.

31.

Turbine Overspeed Trip NA R

NA Mechanism

32. Steam Line Break W

Q R

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Instrumentation And 4

Control System Logic Test 6 Control Circuits 33.

Diesel Generator M

Q NA U

Protective Relaying, Starting Interlocks And Circuitry 34 Off-site Power Undervoltage W

R R

And Protective Relaying Interlocks And Circuitry 35.

Borated Water Storage W

NA R

Tank Level Indicator 36' Boric Acid Mix Tank l

a.

Level Channel NA NA R

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' Temperature Channel M

NA R

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Table 4.1-2 (Continued)

Minimum Equipment Test Frequrag s.

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Item Test Frequency

12. Flow Limiting Annulus

• Verify, at ndrmal One year, two years, on Main Feedwater operating conditions,'

three years, and every Line.s at Reactor that a gap of at least five years thereafter Building Penetration 0.025 inches exists measured from date of between the pipe and initial test.

the annulus.

13. SLBIC Pressure Calibrate Each Refueling Period Sensors

14. Main Steam' Isolation

a. Excercise Through

a. Quarterly Valves Approximately 10%

Travel

b. Cycle

b. Each Refueling Shut-down.

4

15. Main Feedwater

a. Exercisc 'through

a. Quarterly Isolation Valves

.Approximately 5%

Travel 4

b.

Cycle

b. Each Refueling Shut-down.

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