1CAN049606, Application for Amend to License DPR-51,revising as-found Setpoint Tolerance for Pressurizer Code Safety Described by Bases Associated W/Specs 2.2 & 3.1.1 from +1/-3% to +3%

Application for Amend to License DPR-51,revising as-found Setpoint Tolerance for Pressurizer Code Safety Described by Bases Associated W/Specs 2.2 & 3.1.1 from +1/-3% to +3%

ML20108D126
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/29/1996
From:	Dewease J ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20108D129	List: 1CAN049606, Application for Amend to License DPR-51,revising as-found Setpoint Tolerance for Pressurizer Code Safety Described by Bases Associated W/Specs 2.2 & 3.1.1 from +1/-3% to +3% ML20108D136
References
1CAN049606, 1CAN49606, NUDOCS 9605070286
Download: ML20108D126 (4)

Text

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wwn Oreraisur6 Su;4x.xt April 29,1996 1CAN049606 U. S. Nuclear Regulatory Commission Document Control Desk l Mail Station Pl-137 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specification Change Revising The Pressurizer Code Safety As-Found Setpoint Tolerance l

Gentlemen:

l Attached for your review and approval are proposed Technical Specification (TS) changes to allow Arkansas Nuclear One - Unit 1 (ANO-1) to revise the as-found setpoint tolerance for the pressurizer code safeties described by the Bases associated with Specifications 2.2 and 3.1.1 from +1/-3% to i3%. The changes also increase the relief flowrate of the pressurizer code safeties described in the Bases associated with Specification 3.1.1 from 300,000 lb/hr to 324,000 lb/hr, reword the Bases associated with Specification 3.1.7 to describe the actual value of moderator temperature coefficient used as an input to the startup accident analysis, and revise the values for minimum and maximum pressurizer water level specified by Specification 3.1.3.4 to refer to a figure that will be incorporated in this change. These changes are supported by revised startup accident and rod withdrawal accident analyses.

Proposed changes to the ANO-1 Safety Analysis Report incorporating the new analysis results have also been included for your use in reviewing this change request.

The new startup and rod withdrawal accident analyses were performed using the RELAP5/ MOD 2-B&W computer code to justify an increase in pressurizer code safety valve as-found tolerance to +3%. The analyses verified, using conservative assumptior.s, that a

+3% tolerance is acceptable for two pressurizer code safety valves. The analyses also showed that a maximum pressurizer water level of 259 inches below 15% Rated Power and a

maximum level of 320 inches when at or above 15% Rated Power produces acceptable results.

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.r-a U. S. NRC April 29,1996 1CAN049606 Page 2 Currently, when a pressurizer code safety valve setpoint is found to be outside of the +1/-3% j setpoint tolerance, the other pressurizer code safety valve must be tested and the occurrence l must be tracked under the ANO 10CFR50 Appendix B corrective action program. With this change, those occurrences when a pressurizer code safety valve setpoint is found outside of a

+1% setpoint tolerance, but within the proposed +3% setpoint tolerance, would not require testing of the other pressurizer code safety valve and would not require tracking of the corrective action. The change still requires any valve setpoint found to be outside of a 1%

tolerance be returned to within the il% as-left tolerance as currently described in the Bases l associated with Specification 2.2.

Entergy operations currently utilizes the 1980 Edition of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code at ANO-1. Subsection IWV-3512 of this Edition of the Code endorses ASME Performance Test Code (PTC) 25.3-1976 for the testing of safety and relief valves. The pressurizer code safety valves are currently tested in accordance with this standard.

In accordance with 10CFR50.55a(f)(4)(iv), Entergy Operations requests approval to use the 1989 Edition of Section XI of the ASME Code to test the ANO-1 pressurizer code safety valves beginning with testing to be conducted during our next refueling outage which is currently scheduled to commence on September 17, 1996. This Edition, which has been incorporated by reference in 10CFR50.55a(b)(2), endorses ASME/American National .

Standards Institute (ANSI) Operations and Maintenance (OM) Code, Part 10 [OMa-1988 l Addenda to the OM-1987 Edition per 10CFR50.55a(b)(2)(viii)]. This Edition of OM Part 10 endorses OM Part 1 (1987), and allows a 3% tolerance for as-found testing of safety valves.

In adopting the 1989 ASME Code for pressurizer code safety valve testing, Entergy Operations commits to adopt all the related requirements of OM Part 1. The ANO-1 safety analysis was reviewed and determined to be unaffected by this change in testing requirements.

The proposed TS change has been evaluated in accordance with 10CFR50.91(a)(1) using I I

criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinatiom are included in the attached submittal.

Entergy Operations requests that the effective date for this TS change be within 30 days of approval. Although this request is neither exigent nor emergency, your prompt review is requested prior to our next refueling outage.

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U. S. NRC April 29,1996 1CAN049606 Page 3 Very truly yours.

LDn JG cws Attachments i

l To the best of my knowledge and belief, the statements contained in this submittal are j true. 1 SUBSCRIBED AND SWORN TO before me, a Notary Public in and for ML.

County and the State of Mississippi, this 450, day of L Lt ,1996.

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, e -. , a U. S. NRC April 29,1996 1CAN049606 Page 4 cc: Mr. Leonard J. Callan Regional Administrator U. S. Nuclear Regulatory Commission Region IV ,

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident inspector .

Arkansas Nuclear One l

P.O. Box 310 London, AR 72847  ;

Mr. George Kalman

'6RR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission r NRR Mail Stop 13-H-3 i One White Flint North l 11555 Rockville Pike i R.ockville, MD 20852 Mr. Bemard Bevill Acting Director, Division of Radiation Control and Emergency Management -

Arkansas Department of Health 4815 West Markham Street Little Rock, AR 72205 l

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