ML20235K444

Safety Evaluation Supporting Granting Licensee Relief from Volumetric Exam of Steam Generator Primary Side Noozles Inside Radiused Sections

ML20235K444
Person / Time
Site:	Farley
Issue date:	07/08/1987
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235K423	List: ML20235K421 ML20235K444
References
TAC-64435, TAC-64436, NUDOCS 8707160320
Download: ML20235K444 (5)

Text

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...../ I SAFETY EVALUATION FOR GRANTING 0F RELIEF BY fHE OFFICE ,

0F NUCLEAR REACTOR REGULATION RELATED TO VOLUMETRIC EXAMINATION OF STEAM GENERATOR PRIMARY SIDE N0ZZLES ALABAMA POWER COMPANY j

JOSEPH M. FARLEY NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364 I. BACKGROUND INFORMATION The Technical Specifications for the Joseph M. Farley Nuclear Plant Units 1 and 2 state that inservice examination of ASME Code Class 1, 2 and 3 components shall be performed in accordance with the applicable edition and addenda of Section XI as required by 10 CFR 50.55a(g) except where i specific written relief has been granted by the Commission. The I applicable Code edition and addenda for the Joseph M. Farley Nuclear Plant Units 1 and 2, and upon which the ten-year inservice inspection programs are based, are the 1974 Edition and Addenda through the Summer )

1975. Some examination requirements of Section XI are impractical to perform on older plants, and in recognition of this, the regulation allows the licensee to provide information to the Commissior, in support of its determination that a requirement is impractical and to request relief from that requirement. This information is reviewed and evaluated. If necessary findings are made, relief from the requirement l

may be granted. 1 By letter dated January 13, 1987, Alabama Power Company (the licensee) requested relief from the volumetric examination requirement of Section XI, 1974 Edition through Summer 1975 Addenda, for the steam generator primary side nozzle inside radiused sections and provided information in support of the request. The request, Code requirements, supporting information, and the staff's bases for granting the request are given below.

II. RELIEF REQUEST Relief is requested from the volumetric examination of the steam generator primary side nozzle inside radiused sections (Item No. B3.2, Category B-D).

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~2-III. CODE EXAMINATION REQUIREMENT Table IWB-2600, Item B2.2, 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 eactt of the steam generator primary side nozzle-to-shell welds and the adjacent inside radiused sections.

IV. LICENSEE'S 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 coupling and the geometrical configurat. ion requires an excessively long test metal distance resulting in high ultrasonic attenuation. The inside of the nozzle and channel head areas are covered with cladding in the "as-welded" condition; therefore, meaningful volumetric examination cannot 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 metal interface.

1 V. LICENSEE'S PROPOSED ALTERNATIVE EXAMINATION The inside surface of each steam generator primary side nozzle inner radiused section will be visually examined. The examination area will include the inner radius surface region shown in Section XI, Figure IWB-2500 D to the extent practical.

VI. STAFF EVALUATION-AND CONCLUSIONS The design of the steam generator primary nozzles preclude a meaningful volumetric examination of the inner radiused sections being performed.

The as-cast outside surfaces and as-welded clad inside surfaces prevent proper coupling of an ultrasonic transducer. Since there are no welds at the nozzle-to-channel head junctions, the areas were not required to be prepared for ultrasonic examination during the vessel fabrication stage.

Radiation levels prevent radiography of the areas being performed with interpretable results due to film fogging. To impose the requirement tnt the outside or inside surfaces be prepared for ultrasonic examination would create a burden on the licensee without a compensatory gain in safety of the plant.

The licensee has proposed to visually examine the inner radiused areas.

Since the primary purpose of the cladding is to protect the ferritic material from degradation that could result from direct exposure to the reactor coolant environment, the visual examination proposed will provide assurance that clad cracking has not occurred during plant inservice operation. Thus, the cladding will continue to perform its primary function.

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. The staff finds the licensee's proposed alternative visual examination acceptable. The alternative examination will provide assurance of the continued structural integrity of the cladding and of the nozzle-to-channel head junctions. Therefore, we conclude that relief from the volumetric examination requirement for the inner radiused areas of the steam generator nozzles may be granted as requested by the licensee.

DATE: JUL 0 81987 PRINCIPAL CONTRIBUTOR:

G. Johnson I

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2 and regulations. The Commission has made appropriate findings as required I by the Act and the Commission's rules and regulations in 10 CFR Chapter 1.

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A copy of our Safety Evalution for granting of the relief is enclosed.

Sincerely,

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Elinor G. Adensam, Director i Project Directorate 11-1 i

Division of, Reactor Projects I/II '

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Enclosure:

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Mr. R.P. Mcdonald and regulations. The Commission has made appropriate findings as required by the Act and the Commission's rules and regulations in 10 CFR Chapter 1.

A copy of our Safety Evaluation for granting of the relief is enclosed.

Sincerely, Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects I/II

Enclosure:

As stated cc: w/ enclosure:

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