1CAN079901, Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages

Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages

ML20209H573
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/14/1999
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20209H579	List: 1CAN079901, Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages ML20209H581
References
1CAN079901, 1CAN79901, NUDOCS 9907210017
Download: ML20209H573 (6)

Text

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ym Entergy Oper:tions,Inc.

1448 5H 333 RusseMet AR 77801 Tel 50b858-4a88 C. Randy Hutchinson vce Pmsdent OperaDons A,f]

July 14,1999 ICAN079901 U. S. Nuclear Regulatory Commission i Document Control Desk Mail Station OPI-17 Washington, DC 20555 l

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Proposed Technical Specification Changes Revising Reactor Building Structural Integrity Requirements Gentlemen:

Attached for your review and approval are proposed changes to the current Entergy Operations submittal dated April 9,1999 (ICAN049901), regarding the revision of reactor building structural integrity requirements contained in the Arkansas Nuclear One - Unit 1 (ANO-1) Technical Specification (TS). The proposed changes affect ANO-1 TS Administrative Controls and applicable Bases relevant to inservice inspection reporting for the containment structures, tendons, and anchorages.

The proposed changes contained in our April 9,1999 letter are a result of the revisions to 10 ,

CFR 50.55a published in the Federal Register (61 FR 41303), made effective on September 9, 1996, and required to be fully implemented by September 9, 2001. These revised requirements affect the surveillance methods for the containment tendons, the conduct of containment visual inspections, and the reporting methods employed in disseminating the results of these inspections to the Nuclear Regulatory Commission (NRC). ANO-1 is impi,. enting a containment inspection program to comply with the revised requirements of 10 CFR 50.55a which is based on the American Society of Mechanical Engineers (ASME)

Section XI, Subsection IWL. The proposed change is consistent with the current regulatory requirements as described in 10 CFR 50.55a(g)(6)(ii)(B).

Through recent discussions with the NRC, Ente;gy Operations proposes the attached changes to the bases and reporting sections of our April 9,1999 submittal. The last statement of the associated bases has been modified and, in addition, ANO-1 is adding a new reporting requirement to the Administrative Controls section of the ANO-1 TSs relevant to the containment inspection tests results. The format and requirements chosen are consistent with that of the Improved Standard TSs per NUREG-1430 and exceed the requirements set forth in 10 CFR 50.55a and the ASME code.

i 9907210017 990714 anocn osoopy3 m

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U. S. NRC

.. July 14,1999

,1CAN079901 P ge 2 The "No Significant Hazards Determination" contained in our April 9,1999 submittal is unaffected by these proposed changes. Therefore, ANO does not propose changes to the no significant hazards determination as evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c).

The proposed modifications to our April 9,1999 letter provide the evaluation and reporting criteria associated with the inspection results of containment inspection surveillances. Since surveillance of the ANO-1 reactor building will start prior to unit shutdown for refueling outage IRIS, currently scheduled to commence on September 10,1999, Entergy Operations requests NRC approval by August 1,1999, with an implementation period of 30 days.

Very truly yo s, j ////

y f

CRI- dbb Attachment 1

To the best of my knowledge and belief, the statements contained in this submittal are l true.

SUBSCRIBED AND SWORN TO before me a Notary P,9blic in and for County and the State of Arkansas, this /M gday of CC ,1999. '

U 0

" OFFICIAL SEAL" l l 2- __ -- 4.u u_ ) Andrea Pierce J Notary Public, State of Arkansas C Notary Public County of Pope l My Commission Expires /41//S,/,1.60 7 gy Commission Exp.J2n5]2%7]

L U. S. NRC July 14,1999

,1CAN079901 Page 3 I

cc: Mr. Ellis W. Merschoff Regional Administrator U. S. Nuclear Regulatory Commission ,

Region IV {

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 -

l NRC Senior Resident Inspector l Arkansas Nuclear One P.O. Box 310 l

London, AR 72847 i Mr. Nick Hilton NRR Project Manager Region IV/ANO-1 U. S. Nuclear Regulatory Conunission i NRR Mail Stop 13-D-18 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. David D. Snellings Director, Division of Radiation Control and Emergency Management Arkansas Depanment ofHealth 4815 West Markham Street Little Rock, AR 72205 1

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ATTACHMENT TO ICAN079901

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PROPOSED TECHNICAL SPECIFICATION {

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RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.

ARKANSAS NUCLEAR ONE. UNIT ONE DOCKET NO. 50-313 1

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. Attachment to ICAN079901 Page1 of2 DESCRIPTIGN OF PROPOSED CHANGES The proposed changes to the Arkansas Nuclear One, Unit 1 (ANO-1) Technical Specifications (TS) are as follows:

• The bases for Specification 3.6 have been revised to clarify the reactor building inspection evahiation requirements.

• The Specification 6.12.4 h!.s been added to page 146 of the Administrative Controls section of the ANO-1 TS, stating the reponing requirements associated with ANO-1 Containment Inspection Program.

BACKGROUND-On January 7,1994, the NRC published in the Federal Register (59 FR 979) a proposed amendment to its regulation,10 CFR 50, " Domestic Licensing of Production and Utilization Facilities," to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE, and Subsection IWL, of Section XI, Division 1, of the ASME Code. Following the required comment period, the NRC issued the final rule effective on September 9,1996. The final mie,10 CFR 50.554(g)(6)(ii)(B),' required licensees to implement the requirements of Subsection IWE and Subsection IWL of the ASME Code, with specified modifications and limitations, by September 9, 2001. The notice in the Federal Register (61 FR 413030) recognized that the final rule had satisfactorily considered the previous guidance piovided in Regulatory Guide 1.35, " Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments," Revision 3.

ANO submittal dated April 9,1999 (ICAN049901) proposed the removal of special reporting requirements for the tendon surveillances under specification 6.12.5.a. Because 10 CFR 50.5.5a(b)(2)(ix) requires specific containment structural degradations to be reported under the inservice inspection (ISI) summary repon, ANO believed retaining repolting criteria in the TSs would be redundant to regulations and procedures already in place. However, as a result of subsequent discussions with the NRC, the NRC staff requested that these reporting requirements remain in the TSs. Tl erefore, ANO proposes to include a general statement of the requirement to repon evaluateu findings of the containment inspection surveillance under new specification 6.12.4. The addition of specification 6.12.4 is intended to clarify those conditions that must be reported and the time interval for submitting the subsequent report.

Under this new specification, those degradations that exceed the acceptance criteria of 10 CFR 50.55a(b)(2)(ix) or Subsection IWL of the 1992 Edition of the 1992 Addenda of the ASME Code will be reported. The specific acceptance criteria are consistent with 10 CFR 50.55a and Lubsection IWL of the ASME Section XI Code, and will be contained in the ANO-1 Containment Inspection Program.

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1CAN079901 l Page 2,of 2  !

In addition, in order to ensure the user can effectively apply the time requirement associated with specification 6.12.4, modification of the bases to specification 3.6 is proposed, specifying the engineering evaluation duration if acceptance criteria is exceeded. In short, degradations must be evaluated for their impact on overall containment operability within 30 days of the l completion of the containment inspection surveillance. The completion of the engineering j evaluation begins the 60 days time allowance to prepare and submit the reactor building inspection report of specification 6.12.4. These requirements will ensure the desired report is submitted to the NRC within 90 days of the completion of the containment inspection surveillance. The 30 days provided for report preparation and submittal is consistent with the improved standard TSs (ITS).

DISCUSSION OF CIIANGE An additional specification 6.12.4 is proposed to address the reporting requirements for the containment building inspection surveillance. In keeping with the ITSs, a time limit of 30 days {

is proposed which begins upon completion of the engineering evaluation required for degradations found that exceed the established acceptance criteria. The acceptance criteria of  ;

10 CFR 50.55a(b)(2)(ix) and Subsection IWL of the ASME code are maintained within the j proposed ANO-1 Containment inspection Program procedures and are not listed in the {

proposed specification for the sake of redundancy. Included in the proposed, specification, however, are elements that must be included in the reactor building inspection report. These elements have been acquired from Subsection IWL-3310 of the 1992 Edition of the 1992 Addenda of the ASME Code.

l The bases statement of specification 3.6, as submitted under ICAN049901, is being revised to require any resulting engineering evaluation to be performed within 60 days of the completion .

of the containment building inspection surveillance. This modification will allow the 1 engineering evaluation to determine an aggregate assessment of all degradations found during the surveillance, and therefore, the combined effect on reactor building operability. As me.doned previously, the changes to specification 6.12.4 and the bases to specification 3.6 will ensure a report of the condition of the reactor building structure and any corrective actions that were required as a result of the inspection is submitted to the NRC within 90 days.

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