ML20073H756

Responds to NUREG-0966, SER Re D2/D3 Steam Generator Design Mod. Supplemental Info,In Form of Proposed Wording for Forthcoming Change to Tech Specs,Re Eddy Current Testing Program,Encl.Changes to Inservice Insp Recommended

ML20073H756
Person / Time
Site:	Summer
Issue date:	04/13/1983
From:	Dixon O SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20073H761	List: ML20073H756 ML20073H764
References
RTR-NUREG-0966, RTR-NUREG-966 NUDOCS 8304190016
Download: ML20073H756 (3)

Text

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SOUTH CAROLINA ELECTRIC & GAS COMPANY

.OST OFFICE 764 COLUMetA. SOUTH CAROLINA 29218 O. W DixON, JR.

VsCE PnESIDENT uvCtEan o.EnafiO .

April 13, 1983

'Mr. Harold R. Denton, Dire'ctor Office of~ Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket lio. 50/395 Operating License No. NPF-12 NUREG-0966

Reference:

Letter from Mr. O. W. Dixon to Mr. H. R. Denton, " Operating License Condition 2.C.14,"

February 14, 1983 4-

Dear Mr. Denton:

This letter is provided in response to NUREG-0966, " Safety Evaluation Report related to D2/D3 Steam Generator Design l

~

Modification," of March 1983. Based on a review of this NUREG, the following' supplemental information is being provided as input to the Virgil C. SummerSection Nuclear Station plant specific Safety Evaluation Report. 3.8 of the NUREG, regarding Inservice. Inspection and Testing, describes additional. testing and surveillance requirements '

beyond those proposed in the referenced submittal. These additional requirements involve Eddy Current Testing (ECT),

Visual Examination of the Manifold, and Loose Parts Monitoring.

In-view of the NRC concerns and-the position stated in l

NUREG-0966, some changes to the previously proposed inservice inspection program, as described in the referenced letter, are recommended. These changes include the following:

1. The addition of the plant's first inservice inspection, as-defined in the plant Technical Specifications, to the j

Design Review Panel (DRP) Report recommended preheater

section inspection.

l l 2. The ECT of a minimum of 240 tubes total in the preheater i

section of all steam generators during the first and second inservice inspection.

bOf

$"idk'oEohoi G

r

. Harold R. Denton NUREG-0966 April 13, 1983

Page-92

3. -Performing an additional visual-examination of the accessible portions of the manifold internal areas during the secondiECT inspection.

It is felt that the recommendations contained herein are adequate to ensure the quality and integrity of the steam generator modification and that-this inspection / examination program should be separate from that in the unaffected' region of the steam generators.

Attachment 1 to this letter, which is provided as an alternative to the ECT Program described in the subject NUREG, includes proposed. wording for the forthcoming change to the Technical Specifications concerning the ECT Program.

A more detailed scope of the proposed Visual Examination program beyond that described in the referenced letter has not been defined due to the "first of a kind" nature of the examination. As field experience with these examinations increases, a more specific examination scope can be determined. Every effort will be made to perform the examinations in accordance with the ASME Code,Section XI, IWA-2211 Visual Examination VT-lf however, because of limited accessibility, some flexibility is needed in performing these examinations that the Code does not allow.

The data obtained from the ECT inspections and visual examinations outlined above, along with ALARA considerations, will be used to determine the scope and nature of further inspections or examinations deemed appropriate.

In reference to a Loose Parts Monitoring Program to detect a structural failure of the manifold, the plant's installed Digital Metal Impact Monitoring System will be operated and maintained in accordance with-Technical Specifications 3.3.3.10 and 4.3.3.10. Any structural failure of a manifold, which results in loose parts, should be detected by the system transducers located on the associated steam generator.

The feedwater system modification, as described in the referenced submittal, is being installed with the exception of minor hardware and control changes. A simplified diagram of the feedwater modification is attached as Figure 1 and reflects the deletion of the check valve in.the forward flushing line. No other piping changes have been made to the

Harold R. Denton NUREG-0966 April 13, 1983 Page #3 design as described in the referenced submittal. The forward and reverse flushing temperature and flow permissives for the Feedwater Isolation Valve operation will be incorported administratively. Concerning the piping associated'with the feedwater modification, piping break blowdown, as well as pipe whip, jet impingement and reactive forces were basic design considerations. The design will be evaluated to ensure that present FSAR commitments and considerations will not be changed or omitted. The final, as bt '. lt , configuration of the feedwater system modification and its associated operating criteria and limitations will be evaluated in accordance with 10CFR50.59 to determine.that an unreviewed safety question does not exist prior to plant startup.

In addition to information provided in this letter, a separate letter is being submitted for a proposed Technical Specification change to add the containment isolation valves in the reverse flushing lines to Table 3.6-1.

If you have any questions, please advise.

Ve trul yo s, O. W. ixon, f/. -

AMP:RBC:OWD/fjc cc: V. C. Summer T. C. Nichols, Jr./O. W. Dixon, Jr.

E. C. Roberts H. N. Cyrus J. P. O'Reilly Group / General Managers O. S. Bradham R. B. Clary C. A. Price A. R. Koon C. L. Ligon (NSRC)

G. J. Braddick J. C. Miller J. L. Skolds J. B. Knotts, Jr.

NPCF File (Lic./Eng.)

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