ML19254E433

Forwards Revised Response to IE Bulletin 79-13,Revision 1, Cracking in Feedwater Sys Piping. Radiographic Examination of Auxiliary Feedwater & Main Feedwater nozzle-to-pipe Welds Will Be Performed

ML19254E433
Person / Time
Site:	McGuire
Issue date:	10/19/1979
From:	Parker W DUKE POWER CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
NUDOCS 7911010100
Download: ML19254E433 (3)

Text

'I 4

IIP-Q 1DuxE POWER COMPANY g

~

' ' .I -

Powna Bun.orwo 422 Soura Cnuncu Srnrzr, CnAarotra, N. C. asa4a WILLIAM 0, PARMER,Jftg .{ :; T 22 h b '. k v.cc Passiotu? October 19, 1979 rett,-o t: 4.t. 7o.

Svt.se Pacouctiose 373-4083 Mr. James P. O'Reilly, Director U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 Atlanta, GA 30303 Re: McGuire Nuclear Station IE Bulletin 79-13, Revision 1 RII:JPO 50-369

Dear Mr. O'Reilly:

Attached is a revised response to IE Bulletin 79-13, Revision I which was transmitted in your letter of August 30, 1979. Duke Power Company's original response to this bulletin was transmitted to you in my letter of September 19, 1979.

,A Ve truly yours,/ I a4 e - d.A -

William O. Parker, Jr. .

THH/sch Attachment cc: Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 2C355 1283 972 E

g o* S, S

C e t

5 th Anniversary q*

%e"owsR #.l 7911010 @

~

. McGuire Nuclear Station Response to IE Bulletin 79-13, Revision 1

1. a. Radiographic examination of the auxiliary feedwater and main feedwater nozzle-to-piping welds on the four Unit I steam generators will be performed. Examination coverage, sensitivity and evaluation will be in accordance with the requirements of the bulletin subject to the limitations imposed by nozzle design. '>=zle configuration and com-position may preclude the examination stJ.aitivity from meeting the 2T pentrameter requirement.

b. If cracking is identified in the nozzle-to-piping weld of either the auxiliary or main feedwater nozzles, then the welds in the auxiliary or main feedwater lines up to the first piping support or snubber and high stress points in containment will be examined as in la above.

Any unacceptable code discontinuties will be evaluated and either repaired or justified as acceptable.

c. The current inspection program for safety-related piping supports as described in Mr. W. O. Parker's letter to Mr. J. P. O'Reilly of May 11, 1979, pursuant to 10 CFR 50.55e (SD 269-370/79-02) satisfies the visual inspection requirement of this section of the bulletin.

2. a. Not applicable to McGuire steam generators,

b. A volumetric examination will be performed again at the first refueling outage or alternatively, this examination may be conducted at an outage prior to the first refueling but af ter at least one year of operation above 30% power.

c. A visual inspection of feedwater system piping supports and snubbers in containment will be performed concurrent with the examination described in 2b. This inspection will verify operability and con-formance to design.

3. Not applicable to McGuire Nuclear Station.

4. Any cracking or other unacceptable code discontinuities will be reported to Region II within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. a. Examination of the feedwater nozzle-to-piping welds is in progress and is expected to be completed by December 3, 1979. The inspection of safety related piping supports will be complete prior to fuel loading of Unit 1.

b. Duke Power Company is currently revising both the operating and emergency procedures addressing a feedwater line break accident. These procedures are being revised to enhance operator ability to recognize, and recover from this accident.

c. The primary method of detecting feedwater leakage inside containment is by mo.itoring the containment floor and equipment sump level. The sump level detectors can sense a sump level change of one half inch. There-fore e leakage rate of one gallon / minute inside containment is detectable after approximately forty minutes. Other methods of detecting leakage inside containment are discussed in the McGuire FSAR, Section 5.2.7.

Although the primary emphasis of this section concerns reactor coolant system leakage the methodology is appropriate for detecting feedwater leakage. Sampling the fluid in the containment sump accurately iden-tifies the source of the leakage.

1283 J73

~

s 6.

A written report describing the results of the examination in item I will be submitted by January 3,1980. Any corrective measures that are necessary be described in the report. as a result of this examination will 1283 ')74 4