ML20212J766

Requests Relief from ASME Code,Section XI,1974 Edition Through Summer 1975 Addenda,Table IWB-2600,Item B3.2 & Table IWB-2500,Category B-D,permitting Visual Exam of Each Steam Generator Primary Side Nozzle.Fee Paid

ML20212J766
Person / Time
Site:	Farley
Issue date:	01/13/1987
From:	Mcdonald R ALABAMA POWER CO.
To:	Rubenstein L Office of Nuclear Reactor Regulation
References
TAC-64435, TAC-64436, NUDOCS 8701280296
Download: ML20212J766 (4)

Text

r-O orh18 S re t Post Office nom 2641 nirmmgham, Alatsama 352914400 Telephone 20$ 2501835 NabamaPower

$$loNceN.,oen1 January 13, 1987 Docket Nos. 50-348 50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Attention:

Mr. L. S. Rubenstein Gentlemen:

Joseph M. Farley Nuclear Plant - Units 1 and 2 Inservice Inspection Program for ASME Code Class 1, 2 and 3 Components Alabama Power Company has recently completed a design review of the Unit 1 151 Program to assure that all of the inspection requirements for the first ten-year interval have been identified. During this review, it was determined that one additional relief request is reeded whero performance of a required examination is impractical. This relief request is applicable to both Units 1 and 2.

The ASME Code,Section XI, 1974 Edition through the Sumer 1975 Addenda, Table IWB-2600, Item 03.2 and Tablo IWD-2500, Category D-D require that a 1001 volumetric examination be performed on each o' the steam generator primary side nozzle-to-shell wolds and the adjacent insido radiused sections.

The steam generator primary sido nozzles are integrally cast as a part of the channel head and therefore no wolds exist which requiro volumetric examination of this area.

The geometrical configurations of the nozzle exterior area and the outside and inside surfaCo Conditions prohibit a meaningful volumetric examination of the inner radiused section.

In addition, the doso rates inside the steam generator channel head would prohibit preparation for and performance of the examination from inside the nozzle. Based on those considerations Alabama Power Company herewith requests relief from the ASME Code requirements pursuant to 10CFR50.55a(g)(6)(f).

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Director, Nuclear Reactor Regulation January 13, 1987 U. S. Nuclear Regulatory Commission Page 2 As an alternative, Alabama Power Company proposes to perform a visual examination of each steam generator primary side nozzle inner radiused section. The area to be examined will include, to the extent practical, the primary nozzle inner radiused section surface region shown in Figure IWD-2500 D of Section XI.

l summarizes the requested relief.

It is respectfully requested that the NRC grant this relief for both units by July 14, 1987 to support planning for the Unit 1 eighth refueling outage scheduled for March 1988 when all remaining examinations for the first ten-year l

interval must be completed.

As verhally discussed with the NRC on October 23, 1986, Alabama Power Company is deferring the completion of thirty-eight Unit 1 examinations until the eighth refueling outage. This deferral is within the twelve l

i month extension permitted by the ASME Code,Section XI, IWA-2400.

Sixtean Class 1 examinations include the volumetric or visual l

examination of the reactor vessel wolds, interior surfaces and lower internals. The remaining twenty-two examinations consist of the A and 0 loops main steam alping longitudinal welds which require examination in accordance with t1e augmented inspection program as specified in i

Technical Specification 4.4.11.3. provides a detailed l

listing of the deferred examinations.

I Enclosed is the required application fee of $150.00.

If there are any questions or if additional information is needed, please advise.

Respectfully submitted,

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R. P. Mcdonald RPM /ST0 cs1/0393 Attachments cc: Mr. L. D. Long Dr. J. N. Grace Mr. E. A. Reeves Mr. W. 11. Gradford

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ATTACHMENT 1 i

RELIEF REQUEST: Relief is requested from the volumetric examination of r

the steam generator primary side nozzle inside radiused sections (Item i

No. B3.2, Category B-D).

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Code Examination Requirement i

Table IWB-2600, Item 83.2 and Table IWB-2500 Category B-D of the 1974 Edition through the Summer 1975 Addenda of the ASME Code,Section XI, require that a 100% volumetric examination be performed on each of the steam generator primary side nozzle-to-shell welds and the adjacent inside radiused sections.

I Basis for Relief The steam generator primary side nozzles are integrally cast as a part of the channel head; therefore, no welds exist which require volumetric examination. The steam generator nozzle inner radiused section cannot be volumetrically examined from the outside of the nozzle or channel head because the rough, as-cast contact surface is not suitable for ultrasonic r

r coup 1tng and the geometrical configuration requires an excessively long test metal distance resulting in high ultrasonic attenuation. The inside j

of the nozzle and channel head areas are covered with cladding in the i

"as-welded" condition; therefore, meaningful volumetric examination cannot i

be performed from the "as-welded" surface. Even with proper preparation of the inside surface for volumetric examination, an adequate examination of the area of interest (base metal just below the cladding) could not be achieved due to the resulting ultrasonic response at the clad-to-base 1

metal interface.

Alternative Examination The inside surface of each steam generator primary side nozzle inner i

radiused section will be visually examined. The examination area will include the inner radius surface region shown in Section XI, Figure IWD 2500 0 to the extent practical.

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ATTACHMENT 2 UNIT 1 EXAMINATIONS DEFERRED TO THE EIGHTH REFUELING OUTAGE EXAM.

ITEM N0.

CATEGORY QUANTITY DESCRIPTION

)

81.1 B-A 6

Reactor Vessel Beltline Welds B1.2 B-B 1

Reactor Vessel Lower Head-to-Shell Weld i

i Bl.3 B-C 1

Reactor Vessel Flange-to-Shell Weld B1.4 B-D 3

Reactor Vessel Inlet Nozzle-to-Shell Welds and Nozzle Inside Radiused i

Sections Bl.6 B-F 3

Reactor Vessel Inlet Nozzle-to-Safe End l

Welds i

B1.15 B-N-1 1

Reactor Vessel Interior i

Bl.17 B-N-3 1

Core Support Structure (Lower Internals)

I N/A C-G 11 Loop A Main Steam Piping Longitudinal 1

Welds i

N/A C-G 11 Loop B Main Steam Piping Longitudinal j

Welds L

i i

l j

I l

i

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