ML17334B555

Requests Relief from Requirements of 10CFR50.55 Due to Inability to Meet Requirements Specified in 10CFR50.55. Drawings Encl

ML17334B555
Person / Time
Site:	Cook
Issue date:	07/28/1995
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9508030164
Download: ML17334B555 (17)

Text

~ PRJORXTY 1i (ACCELERATED RZDS PROCESSING)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION -NBR:9508030164 DOC.DATE: 95/07/28 NOTARIZED: NO DOCKET g FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 p 50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Requests relief, from requirements of 10CFR50.55 due to inability to meet requirements specified in 10CFR50.55.

Drawings encl.

DISTRIBUTION CODE: 'A001D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 N,J 1 1 INTERNAL: PB~GENTE ' 1 1 NRR/DE/EMCB 1 1 NRR/DRCH/HICB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS2 1 0 EXTERNAL: NOAC 1 1 NRC PDR 1 1 D N

NOTE TO ALL "RZDS" RECZPZENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 11

Inrliana Michigan Power Company P.O. Box 16631 Columbus, OH 43216 IHEMSIHA Cmari:mamas i'mWM July 28, 1995 AEP:NRC'0969AI Docket Nos.: 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:

Donald C. Cook Nuclear Plant Units 1 and 2 REQUEST FOR RELIEF FOR AUGMENTED REACTOR VESSEL IN-SERVICE INSPECTION The purpose of this letter is to request relief from the requirements of 10 CFR 50.55a(g)(6)(ii)(A)(5) (augmented reactor vessel shell weld inspection) because we have found that we are unable to meet the requirements specified in 10 CFR 50.55a(g)(6)(ii)(A). Specifically, we are unable to examine greater than 90% of seven reactor vessel welds, because of interferences created by structures within the reactor vessel (see attachment).

The amount of weld accessibility for inspection on the unit 1 reactor vessel is provided in the attachment to this letter. The amount of weld accessibilities on the unit 2 reactor vessel is expected to be approximately the same as unit 1. The ASME Code,Section XI requires that essentially 100% of the weld be examined, and, for the augmented inspection, 10 CFR 50.55a(g)(6)(ii)(A)(2) defines "essentially" as greater than 90%. We are, therefore, requesting code relief for this inspection. The bases for the request are presented in the attachment to this letter. This reactor vessel weld inaccessibility condition is also being reviewed by the Nuclear Regulatory Commission for other plants, such as Ft. Calhoun.

The augmented reactor vessel weld inspection is required to be completed by the end of our second ten-year inspection interval, June 30, 1996. We intend to perform these examinations during the next refueling outages for unit 1 and unit 2, currently scheduled for September 16, 1995, and March 16, 1996, respectively.

0~(,MQQ 9508030ih4 950728 o4, PDR ADOCK 050003i5

..PDR.

U. S. Nuclear Regulatory Commission AEP:NRC:0969AI Page 2 Approval of the relief requests is not required until the end of the respective ten year intervals, June, 1996 for both unit 1 and unit 2. An expeditious review, however, would be beneficial in view of the upcoming outages.

Sincerely, E. E. Fitzpatrick Vice President pit Attachment cc: A. A. Blind G. Charnoff H. J. Miller NFEM Section Chief NRC Resident Inspector - Bridgman J. R. Padgett

Attachment to AEP:NRC:0969AI Background Information and Justification 10 CFR 50.55 Code Relief For the Augmented Reactor Pressure Vessel Shell Welds Examination for Cook Nuclear Plant Units 1 & 2

Attachment to AEP:NRC:0969AI Page 1 Background for Augmented Vessel Examination Code Relief Request I Code Relief Request Code relief is requested for the following reactor pressure vessel (RPV) shell welds which are scheduled to be inspected during the next units 1 & 2 refueling outages which will be the final outages for the second ISI interval:

Category I.D. Item ¹ Component description urine B-A Bl.ll Circumferential weld (lower head to shell weld)

B-A B1.12 Longitudinal shell weld (upper shell at 26.5 degrees)

B-A B1.12 Longitudinal shell weld (upper she11 at 146.5 degrees)

B-A B1.11 Circumferential Veld (lower head to shell weld)

B-A B1.12 Longitudinal shell weld (upper shell at 22 degrees)

B-A B1.12 Longitudinal shell weld (upper shell at 113 degrees)

B-A B1.12 Longitudinal shell weld (upper shell at 247 degrees)

II Code Requirements ASME Section XI, 1983 Edition Summer Addendum, Table IWB-2500-1, Category B-A, Item B1.10 requires volumetric examination of one beltline region of the RPV shell welds for each ten year interval following the first ten year interval.

10 CFR 50.55a(g)(6)(ii)(A) requires that an augmented reactor vessel weld inspection be conducted prior to the end of the current interval. 10 CFR 50.55a further states that essentially 100% of the weld length (no less than 90%) is to be examined in and if lesswiththan10 90%, code relief must be requested accordance CFR 50.55a(g)(6)(ii)(A)(5) with the justification for doing less than 90%.

Attachment to AEP:NRC:0969AX Page 2 III Basis for code relief Figures 1 & 2 identify the RPV shell welds for Cook Nuclear Plant units 1 & 2 RPVs. Table 1 identifies the welds for which relief is requested and indicates the estimated examination coverage percentages for unit 1. The unit 2 coverage is expected to be similar to the unit 1 coverage.

Reactor pressure vessel shell welds are examined from the inside surface using automated ultrasonic equipment. The examination of the shell to lower head weld is limited to less than 90% due to the position of the core support lugs which provide an anti-rotation feature for the core barrel.

These core support lugs inhibit the equipment access required to perform a code ultrasonic (UT) exam of the shell weld from both sides of the weld.

We also anticipate that the longitudinal upper shell welds which intersect outlet nozzles can not be examined at coverage percentages of 908 or better due to physical and geometric interferences (see Table 1).

The automated examinations on both units 1 & 2, which are scheduled for the next refueling outages, (unit 1 1995 and unit 2 1996), will be performed with modified equipment and tooling designed to accommodate the maximum coverage possible. Automated equipment set-up will also be optimized (indexed as close to the obstructions as possible) to afford maximum coverage.

Alternate examination from the outside surface of the lower shell weld to the RPV is not practical due to high levels of radiation expected for the personnel who will install and tear down 'scaffolding and insulation, and for the personnel who will conduct the volumetric examinations.

IV Alternate Examinations No alternate examinations are proposed at this time. Special tooling has been designed and will be used to maximize coverage. Equipment set-up (indexing as close to the obstructions as possible) will be used during the upcoming outages.

Justification for Granting of Code Relief Examination of 100 percent of RPV shell welds is not practical. Examination of the accessible weld volume should be sufficient to provide reasonable assurance of vessel integrity. This change will not endanger life or property or the common defense and security because the reactor coolant

Attachment to AEP:NRC:0969AI Page 3 system is designed and constructed to have a low probability of gross rupture or significant leakage throughout its design life and technical specification 3.4.6.2 places limits on the amount of reactor coolant system leakage during operation.

The most likely weld failure would be a crack which would allow reactor coolant to leak from the system. Any such leakage would be detected and retained within the containment building. Should this occur, the appropriate action statement would be followed if the leakage exceeded the Additionally, past technical specification allowables.

examinations of the accessible RPV shell welds have revealed no recordable indications and it is reasonable to conclude the same results for these inaccessible welds would be obtained.

Alternate examination from the outside of the RPV is impractical due to the expected high radiation exposure associated with the scaffolding and insulation removal and replacement and UT examination with no commensurate safety benefit realized.

4 Table 1 Relief Request Estimated Examination Coverages For Cook Nuclear Plant Units 1 & 2 Unit Veld Exam Area Estimated Comments Number Xdentification Coverage (4)

RPV-D Lower shell to 76 Limitation due to core lower head. support anti-rotation lu s.

RPV-VA1 RPV-VA2 RPV-D

. Upper 26.5' Upper 146 Lower 5'8 shell at shell at shell to lover head.

85 Unknown+

Limitation due to intersectin nozzle.

Limitation due to intersectin nozzle.

Limitation due to core support anti-rotation lu s.

RPV-VA1 Upper shell at Unknown+ Limitation due to 22'. intersectin nozzle.

RPV-VA2 113'nknown+

Upper shell at Limitation due to intersectin nozzle.

RPV-VA3 247'nknown+

Upper shell at Limitation due to intersectin nozzle.

• Unit 2 percentage coverage estimate is not available but is believed to be approximately 768 for the lower shell. to lower head weld and 85-88%

for the intersecting nozzles.

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