ML19289F537

Forwards Proposed Changes to Tech Specs Limiting Conditions for Operation Related to Anticipatory Reactor Trip on Loss of Main Feedwater &/Or Turbine Trip

ML19289F537
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/06/1979
From:	Cavanaugh W ARKANSAS POWER & LIGHT CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
1-069-2, 1-69-2, NUDOCS 7906110145
Download: ML19289F537 (9)

Text

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ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 UTTLE ROCK. ARKANSAS 72203 (501) 371-4422 WILLIAM CAVANAUGH lli June 6, 1979 Vice President Generation & Constructaon 1-069-2 Director of Nuclear Reactor Regulation ATTN: Mr. Robert W. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Arkansas Nuclear One-Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specification (File: 1511.1)

Gentlemen:

Attached are proposed changes to the ANO-1 Technical Specifi-cations. The proposed changes add limiting conditions for operation related to the anticipatory reactor trip on loss of Main Feedwater and/or Turbine Trip. Proposed surveillance requirements related to the anticipatory reactor trips and actuation of the emergency feedwater pumps are also included.

The bases for these changes are as follows.

Our proposed specification 3.5.1.7 requires that the antici-patory reactor trip circuitry on loss of Main Feedwater and/or Turbine Trip be operable above the applicable reactor power levels as defined in the May 17, 1979, Order SER.

Our proposed Specification 3.5.1.8 is identical to the current specification 3.5.1.7. We propose that it be renumbered to provide greater continuity in our specifications.

The proposed surveillance requirements related to the antici-patory trips and emergency feedwater actuation have been added to Table 4.1-1 of the specifications. These changes are re-quired by the May 17, 1979, Order SER and will ensure the continuing reliability of the subject circuitry.

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1-069-2 Mr. R. W. Reid June 6, 1979 We have determined that the proposed change is a Class III Amendment in accordance with 10 CFR 170.22 and accordingly have submitted a check for $4,000.00.

Proposed changes to specifications related to the emergency feedwater system will be submitted June 8, 1979 as dis-cussed with the staff.

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l' on William Cava ugh, II WC/M0W/ew Attachment 2230 321

STATE OF ARKANSAS )

) SS COUNTY OF PULASKI )

William Cavanaugh III, being duly sworn, states that he is Vice President, Generation & Construction, for Arkansas Power & Light Company; that he is authorized on the part of said Company to sign and file with the Nuclear Regulatory Commission this Supplementary Information; that he has reviewed or caused to have reviewed all of the statements contained in such information, and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, informa-tion and belief.

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'^ y William Cav ugh I SUBSCRIBED AND SWORN T0 before me, a Notary Public in and for the County and State above named, this h ay of /M 1979.

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WAze Notary Publit / [

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My Conmission Expires:

My Commission Expires 9/1/81

3.5.1.7 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 10% reactor power.

2. Reactor trip upon Turbine Trip shall be operable (as determined by Specification 4.1.a, items 2 and 42) at greater than 20% reactor power.

3. If the requirements of Specifications 3.5.1.7.1 or 3.5.1.7.2 cannot be met, restore the inoperable trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or reduce reactor power to below the applicable limit at the maximum safe rate.

3.5.1.8 The Decay Heat Removal system isolation valve closure setpoints l shall be equal 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 DHR system shall be equal to or less than 450 psig.

2230 323 42a

Table h.1-1 Instrument Surveillance Pequirements Channel Description Check Test Calibrate Remarks

1. Protective Channel NA M UA Coincidence Logic

2. Control Rod Drive NA M NA (1) To include tripping of breakers via shunt Trip Breaker trip circuit.

3 Power Range Amplifier NA NA T/W(1) (1) Heat balance calibration twice weekly under steady state operating conditions, daily under non-steady state operating conditions.

4. Power Range Channel S M M(1)(2) (1) Using incore instrumentation.

M(1) (2) Axial offset upper and lover chambers monthly and after each startup if not done previous week.

S 5 Intermediate Range Channel S P/M NA

6. Source Range Channel S(1) P NA (1) When in service.

7 Reactor Coolant Tempera- S M R ture Channel

8. High Reactor Coolant S M R Pressure Channel 9 Low Reactor Coolant S M R ps, N Pressure Channel (W

CC) 10. Flux-Reactor Coolant Flow S M R Comparator 6

U\3 11. Reactor Coolant Pressure S M R 4** Temperature Comparator

12. Pump Flux Comparator S M R

D Table h.1-1 (Cont'd) csJ A

Channel Description Check Test Calibrate Remarks cc)

13. High Reactor Building D S M R CNJ Pressure Channel CNJ lb. High Pressure Injection NA M NA Logic Channel 15 High Pressure Injection Analog Channels

a. Reactor Coolant S M (1) R (1) Including test of shutdown bypass Pressure Channel function (ECCS bypass function).

b. Reactor Building S M R h psig Channel w

O 16. Low Pressure Injection NA M NA Logic Channel 17 Low Pressure Injection Analog Channels

a. Reactor Coolant S M (1) R (1) Including test of shutdown bypass Pressure Channel function (ECCS bypass function).

b. Reactor Building S M R h psig Channel

18. Reactor Building Emergency NA M NA Cooling and Isolation System Logic Channel 19 Reactor Building Emergency Cooling and Isolation System Analog Channels

a. Reactor Building S M R h psig Channels L6

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N N.

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Tab l e .4_.1 - 1 (Cont'd) r, N Channel Description Check Test Calibrate Remarks

30. Decay IIcat Removal S(l)(2) M(1)(3) R (1) Includes RCS Pressure Analog g; System Isolation Valve Channel Automatic Closure and Interlock System (2) includes CFT Isolation Valve Position

. (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 R$ 32. Steam Line Break W Q R Instrumentation And Control System Logic Test 6 Control Circuits

33. Diesel Generator M Q NA 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 C) Tank Level Indicator (je 36. Reactor Trip Upon Loss of Main NA PC NA py Feedwater Circuitry Ch

s

. Table 4.1-1 (Cont'd)

N

@ Channel Description Check Test Calibrate Remarks a

@ 37. Boric Acid Addition Tank f a. Level Channel NA NA R

_El b. Temperature Channel M NA R

$ 38. Deleted

39. Sodium Hydroxide Tank NA NA R Level Indicator

40. Incore Neutron Detectors M(1) NA NA (1) Check Functioning

41. Emergency Plant Radiation M(1) NA R (1) Battery Check u Instruments Y

42. Reactor Trip Upon NA PC NA Turbine Trip Circuitry

43. Strong Motion Acceleographs Q(1) NA Q (1) Battery Check

44. ESAS Manual Trip Functions

a. Switches & Logic NA P NA

b. Logic NA M NA p) 45. Reactor Manual Trip NA P NA N

(ja 46. Reactor Building Sump Level NA NA R CD u

N

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[ Tp_ble 4.1-1 (Cont'd)

S.

M S Channel Description Check Test Calibrate Remarks z

? 47. EFW Actuation Control Logic NA NA R

, Note: S - Each Shift T/W - Twice per Week R - Once every 18 months D - Daily B/M - Every 2 Months NA - Not Applicable W - Weekly Q - Quarterly PC - Prior to Going Critical if Not Done Within Previous 31 Days M - Monthly P - Prior to Each Startup if Not Done Previous Week U

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