ML18025A274

Letter Responding to the Commission Letter of August 9, 1976 Relating to Annulus Pressurization and Cracks in the Feedwater Nozzle Blend Radii

ML18025A274
Person / Time
Site:	Susquehanna
Issue date:	09/09/1976
From:	Curtis N Pennsylvania Power & Light Co
To:	Parr O Office of Nuclear Reactor Regulation
References
Download: ML18025A274 (6)

Text

U.II, NUCLEAR AEOULATOAVC NIIC FOAM 196 (2. 70 I NRC DISTRIBUTION FoA PART 50 DOCKET MATERIAL SB ION OOOKIQllGMll p

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DATE OF DOCUMENT DATg RECEIVED 7-/3-c LETTER 8onIDINAL QCOr V C3NOTOAIZED QtffVCLASSIF IF. D PROP INPUT FORM NUMBER OF COPIES RECEIVEDtoAI'r DESCRIPTION LTR REF OUR 8-9-76 LTR. ~...FURNXSHING RESPONS TO A QUESTION ON ANNULUS PRESSURIZATION AS THAT INFO WILL 3E A PART OF THE FSAR REVIEW AND FURTHER RESPONDING TO QUET QUESTXON CONCKINING CRACKS IN THE FEEDWATER NOZZLE KEND RANIr, ~

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pter,lL TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 PHONE< t215) 821-5151 Bicentennial htlE19EO P 1S."g

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~ @gag Director of Nuclear Reactor Reg Light Water Reactors Branch No. 3 U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention:

Olan D. Parr, Chief

!g8gulateTIt PZ!(Ei Pil8 SUUEHANNA STEAM ELECTRIC STATION

RESPONSE

TO LETTER DATED 8/9/76 ER 100450 FILE 840-2 PIA-1

Dear Mr. Parr:

Pennsylvania Power 8o Light Company has the following xesponse to your letter of August 9~ 1976:

1.

A response to the question on annulus pressurization will be provided. in conjunction with the FSAR effort.

We intend. to review our methodology with you as part of the FSAR review.

2.

In response to the question concerning the cracks in the feedwater,nozzle blend radii, we have the following response:

General Electric has undertaken an extensive test program to investigate the thermal cycling and, associated nozzle blend. radii cracks which have

occurred, on reactor vessel feedwater nozzles.

The program includes facilities whi!ch can reproduce the thermal cycling phenomena which will allow evaluation of proposed modifications to ensure thermal cycling is reduced..

Testing is scheduled for completion in mid 1977.

Based.

on current test data and, evaluations, a welded thermal sleeve appears to offer the most positive means for correcting the problem.

Additionally, in-service inspection of the nozzle inside radius section of the RPV feedwater inlet nozzles will be accomplished, by ultrasonic examination from the exterior surfaces in accordance with ASME code requirements.

PENNSYLVANIA POWER 8

LIGHT COMPANY

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Director of Nuclear Reactor Regulation Page 2 Tf the investigation of this matter indicates that additional protective measures are warranted, design modifications and/or supplemental in-service inspections will be implemented as appropriate.

Very tru+ yours, N.

W. Curti.s Vice President-Engineering 8c Construction CTC:CAW

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