ML20038A751

Amend 61 to License DPR-51,adding Requirements for Anticipatory Reactor Trip Sys on Loss of Main Feedwater &/Or Turbine Trip

ML20038A751
Person / Time
Site:	Arkansas Nuclear
Issue date:	11/03/1981
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Arkansas Power & Light Co
Shared Package
ML20038A752	List: ML20038A751 ML20038A754 ML20038A759
References
DPR-51-A-061 NUDOCS 8111160385
Download: ML20038A751 (2)

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UNITED STATES

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1 NUCLEAR REGULATORY COMMis5 ION

$..( $ ) f WASHINGTON, D. C. 20555

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ARKANSAS POWER & LIGHT COMPANY

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DOCKET NO. 50-313 ARKANSAS NUCLEAR ONE. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 61 License No. DPR-51 1.

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

A.

The app?ication for amendment by Arkansas Power and Light Company (the licensee) dated June 6,1979, as supplemented February 12, 1980, January 29, 1931, and tiay 29, 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 conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 issuanca of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and I

i E.

The issuance of this amendment is in accordance with 10 CFR Part 1

51 of the Comission's regulations and all applicable requirements l

have been satisfied.

8111160385 811303 DR ADOCK 05000313 p

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

'(2) Technical Specifications

'sThe Technical Specifications contained in-Appendices A and B, as revised through Amendment No. 6% are hereby incorporated in the license. The licensee i

.shall operate the facility in accordance with the' Technical Specifications.,,

3.

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

E0R THE NUCLEAR REGULATORY COMMISSION

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u John F. Stolz, C'hief Op rating Reactors Branch da vision of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 3, 1981 f

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l, ATTACHMENT TO LICENSE AMENDMENT NO. 61

. FACILITY OPERATING LICENSE NO. DPR-51 DOCKET.N0. 50-313

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. P.eplace. the.following pages of the Appendix' "A" Technical Specifications-

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'r. umber and contain vertical linas indicating the area of ch.inge.

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3.5.1.7 The Decay Heat Removal Systen: isolation valve closure setpoints shall be ecual to or less than 340 psig for one valve and equal to or less than 400 psig for the second valve in the suction line. The relief valve setting for the DHR system shall be equal to or less than 450 psig.

3.5.1.8 The degraded voltage monitoring relay settings shall be as follows:

a.

The 4.16 KV emergency bus undervoltage relay setpoints snall be > 3115 VAC but < 3177 VAC.

b.

The 460 V e ergency bus undervoltage relay setpofits shall be > 423 VAC but < 431 VAC with a time delay setpoint of 8 seconds +1 ser.o.1d.

3.5.1.9 The following Reactor Trip circuitry shall be operable as indicated:

1.

Reactor trip upon loss of Main Feedwater shall be operable (as determined by Specification 4.1.a. items 2 and 36 of Table 4.1-2) at greater than 5% reactor power.

(May be bypassed up to 10% reactor power.)

2.

Reactor trip upon Turbine Trip shall be operable (as detennined by Specification 4.1.a. item; 2 and 42) at greater than 5% reactor power.

(May be bypassed up to 20 reactor power.)

3.

If the requirements of Specifications 3.5.1.9.1 or 3.5.1.9.2 cannot be met, restore the it. operable trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or bring the plant to a hot shutdown condition.

Amendment No. JE(61 42a-1

for protective action from a digital ESAS subsystem will not cause that subsystem to trip. The fact that a module has been removed will be con-tinuously annunciated to the operator. The redundant digital subsystem is still sufficient to indicate complete ESAS action.

The testu.s schemes of both the RPS and the ESAS enable complete system testing while the reactor is operating. Each channel is capable of being tested independently so that operation of individual channels may be eval-

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usted.

Reactor trips on loss of all main feedwater and on turbine trips will sense the start of a loss of OT5G heat sink and actuate earlier than other trip signals. This early actuation will provide a lower peak RC pressure dur-ing the initial over pressurization following a 1:ss cf feedwater or tur-bine trip event. The LOF4 trip may be bypassed up to 10% to allow suf-ficient margin for bringing the MFW pumps into use.at approximately 75.

The Turbine Trip trip may be bypassed up to 20% to allow sufficient margin for bringing the turbine on line at approximately 151.

The Automatic Closure and Isolation System (ACI) is designed to close the Decay Heat Removal System (OHRS) return line isolation valves when the Reactor Coolant System (RCS) pressure exceeds a selected fraction of the OHRS design pres?ure or when core flooding system isolation valves are opened. The ACI is designed to permit manuti operation of the DHRS re-turn line isolation valves when permissive conditions exist.

In addition, the ACI is designed to disallow manual oseration of the valves when per-missive conditions do not exist.

Power is normally supplied to the control rod drive mechanisms from two separate parallel 480 volt sourc6s.

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 Instrumentation and Control System (SLBIC) is designed to automatically close the Main Steam Block valves and the Main Feedwater Isolation valves upon loss of pressure in either of the two main steam lines.

The SLBIC is also designed to be reset from its trip position only when the system is shut down or the Main Steam line pressure is below 650 psig.

The Degraded Voltage Monitoring relay settings are based on the short term starting voltage protection as well as long term running voltage protection.

The 4.16 KV undervoltage relay setpoints are based on the allowable starting voltage plus maximum system voltage drops to the motor terminals, which allows approximately 78% of motor rated voltage at the motor terminals. The 460V undervoltage relay setpoint is based on long term motor voltage re-quirements plus the maximum feeder voltage drop allowance resulting in a 92% setting of motor rated voltage.

Amendment flo. JCT, M 61 43a

1 The OPERABILITY of the accident monitering instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97. " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During.

and Following an Accident." December 1975 and NUREG-0578. "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations."

REFERENCE FSAR. Section 7.1 P

Amendment tio. 61 43b

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.E Table 3.5.1-1 (Cont'd)

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REACTOR PROTECTION SYSTEM (Cont'd) m 1

2 3

4 5

i No. of Operator action.

clunnels Min.

Min.

if conditions of No. of for sys-operable degree of column 3 or 4 functional Unit channels tem trip channels rededancy cannot be met i

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11. Reactor trip upon loss of N in feedwater' 4

2 2

1 Notes 1, 15

12. Reactor trip upon turbine trip 4

2 2

1 h tes 1. 16 i.

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Table 3.5.1-1 (Cont'd)

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Notes Cont'd.

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13.

Channels muy be bypassed for not greater than 30 seconds during reactor coolant pump s ta rts.

If the automatic bypass circuit or its alana circuit is inoperable, the undervoltage protection shall be restored within I hour, otherwise Note 14 applies.

14.

With the number of channels less-than requited, restore the inoperable channels to opdrable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

15.

This trip function may be bypassed at up to 10% reactor power.

16.

This trip function may be bypassed at up to 20% reactor power.

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g Tahic 4.1-1 (Cont'd) a R

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CleainicI IMscription Check Test Calibrate O

30. Decay lleat Removal S(1)(2)

M(1)(3)

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.k System Isolatinn Valve (1) Includes RCS Pressure Analog Channel g

Autoenat(c Closure and Interlock Systa" (2) Includes CFT Isolation Valve Position y-(3) Shall Also Be Tested During Refuel-ing Shutdown Prior to Repressuriza-tion at a pressure greater than 300 but less than 420 psig.

31. Turbine Overspeed Trip M

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Mecleanism

32. Ste.wi 1.Inc Itreak W

Q R

d lustremientattoes and Coeitrul System logic Test & Control Circults i

33.118esel Generator M

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Protective Helaying Startioni laterlocks And Circuitry

34. Of f-stte Power thodervoltage W

R (1)

R (1)

(1) Shall be tested during refueling shutdown And Protective Relaying Interlocks And Circuitry to demonstrate selective load shedding interlocks function during manual or auto-matic transfer of Unit I auxiliary loads to Startup Transformer No. 2.

35. Borated Water Storage W

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Tank Level Indicator

36. Reactor Trip Upon Loss M

PC R

of Main i~eedwater Cir-cuttry

7 Table'4.1-1 (Cont'd)

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Channel Description Check

' Test Calibrate R. marks

37.. Boric Acid' Addition Tank

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a.. Level Channel NA NA R

b.

Tenperature Channel M

NA R

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38.. Degraded Voltage Monitoring W

R R

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' 39. ' Sodium Hydroxide Tank NA NA R

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Level Indicator 40.

Incore Neutron Detectors M(1)

NA NA (1) Check Functioning

'41.

Emergency Plant-Radiation M(1)

NA R

(1) Battery. Check Instruments

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42. Readtor Trip Upon M

PC R

Turbine Trip Circuitry.

43. Strong M:, tion Acceleographs Q(1)

NA Q

(1) Battery Check

44. ESAS Manual Trip. Functions a.'

Switches & Logic NA P

NA b.

Logic NA M

NA

45. Reactor Manual Trip NA P

NA

46. Reactor Building Sump Level NA NA R

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