ML17334B555
| 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.
(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 ID CODE/NAME PD3-1 LA N,J INTERNAL:
PB~GENTE NRR/DRCH/HICB NRR/DSSA/SRXB OGC/HDS2 EXTERNAL: NOAC COPIES LTTR ENCL 1
1 1
1 1
1 1
1 1
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1 RECIPIENT ID CODE/NAME PD3-1 PD NRR/DE/EMCB NRR/DSSA/SPLB NUDOCS-ABSTRACT NRC PDR COPIES LTTR ENCL 1
1 1
1 1
1 1
1 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 PDR ADOCK 050003i5
..PDR.
o4,
U.
S. Nuclear Regulatory Commission Page 2
AEP:NRC:0969AI 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 B-A B1.12 Longitudinal shell weld (upper shell at 26.5 degrees)
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 and if less than 90%,
code relief must be requested in accordance with 10 CFR 50.55a(g)(6)(ii)(A)(5) with the justification for doing less than 90%.
Attachment to AEP:NRC:0969AX III Basis for code relief Page 2
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 tooling has coverage.
obstructions outages.
examinations are proposed at this time.
Special been designed and will be used to maximize Equipment set-up (indexing as close to the as possible) will be used during the upcoming 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 willnot 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 technical specification allowables.
Additionally, past 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 Number RPV-D Exam Area Xdentification Lower shell to lower head.
Estimated Coverage (4) 76 Comments Limitation due to core support anti-rotation lu s.
RPV-VA1 Upper shell at 26.5' 85 Limitation due to intersectin nozzle.
RPV-VA2 Upper shell at 1465'8 Limitation due to intersectin nozzle.
RPV-D Lower shell to lover head.
Unknown+
Limitation due to core support anti-rotation lu s.
RPV-VA1 Upper shell at 22'.
Unknown+
Limitation due to intersectin nozzle.
RPV-VA2 Upper shell at 113'nknown+
Limitation due to intersectin nozzle.
RPV-VA3 Upper shell at 247'nknown+
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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