ML17328A187
| ML17328A187 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/06/1989 |
| From: | Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Davis A NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| AEP:NRC:1090H, NUDOCS 8910120095 | |
| Download: ML17328A187 (13) | |
Text
ACCKMRATED DlSTKBUTlON DEMONSTRA,TI04 SYSTEM
'EGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) 0 CCESSION NBR:8910120095 DOC.DATE: 89/10/06 NOTARIZED: NO DOCKET FACIL:50-315 Donald C.
Cook Nuclear Power Plant, Unit 1, Indiana 05000315 50-316 Donald C.
Cook Nuclear Power Plant, Unit 2, Indiana
& '05000316 AUTH.NAME AUTHOR AFFILIATION ALEXICH,M.P.
(formerly Indiana
& Michigan Ele RECIP.NAME RECIPIENT AFFILIATION DAVIS,A.B.
Document Control Branch (Document Control Desk)
SUBJECT:
Responds to NRC 890905 ltr re violations noted in Insp Repts 50-315/89-22
& 50-316/89-22.
DISTRIBUTION CODE:
IE01D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Vio ation Response NOTES:
RECIPIENT ID CODE/NAME PD3-1 PD INTERNAL AEOD AEOD/TPAD NRR SHANKMAN,S NRR/DLPQ/PEB NRR/DREP/EPB 10 NRR/PMAS/ILRB12 OE N/J G
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1 RECIPIENT ID CODE/NAME GIITTER,J.
AEOD/DEIIB DEDRO NRR/DEST DIR NRR/DOEA DIR 11 NRR/DREP/RPB 10-NUDOCS-ABSTRACT OGC/HDS 1 RES MORISSEAU / D NRC PDR COPIES LTTR ENCL 1
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Indiana Michigan Power Company P.O. Box 1663)
Coliinibaa Oi 4 '2 i&
Z lNWANA NlCHIGAN POMfFR AEP 'RC '090H Donald C.
Cook Nuclear Plant Units 1 and 2
Docket Nos.
50-315.and 50-316 License Nos.
DPR-58 and DPR-74 INSPECTION REPORTS 50-315/89022 (DRS)
AND 50-316/89022 (DRS);
RESPONSE
TO VIOLATIONS U, S. Nuclear Regulatory Commission Attn:
Document Control Desk Washington, D.
C.
20555 Attn:
A. B. Davis October 6,
1989 Dear Mr. Davis This letter is in response to H. J, Miller's letter dated September 5,
1989, which forwarded the report of the routine safety inspection conducted from July 10 through August 15, 19&9, on activities at the Donald C.
Cook Nuclear Plant Units 1 and 2.
The Notice of Violation attached to Mr. Miller's letter identified two severity level IV violations in the areas of procedural compliance and design control.
These violations are addressed in the attachment to this letter.
Mr. Miller's letter requested that we provide a discussion of our plan of action to ensure that fillet welds installed on socket fittings (socket welds) in small bore piping systems (less than 2
1/2 inches diameter) are of adequate size
~
This letter also requested a response to a perceived ineffectiveness of interim corrective actions implemented in association with our design control process.
These issues are addressed below.
Item 1 also constitutes our special report on socket weld inspections committed to in discussions with your staff.
1.
Verification of Socket Weld Size Acce tabilit In a previous Region III inspection (Inspection Reports 50-315/88028 and 50-316/88032) deficiencies in the documentation of QC inspection and acceptance criteria for socket welds were identified.
The noted deficiencies centered on the weld size specified for original design versus that specified for QC inspections.
In addition, the 891Oi2OO95 89iOOI=
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Mr. A. B. Davis AEP:NRC:1090H actual measured weld size was not consistently documented following QC inspections.
Mr. Miller's letter requested that we provide a plan of action to ensure that adequate weld size was installed during previous modification work.
Cook Nuclear Plant QC/NDE personnel have performed an inspection of a sample of socket welds to permit an assessment of the acceptability of socket weld sizes as installed.
This inspection was initiated prior to receipt of Mr. Miller's September 5 letter.
The acceptance criteria are based on the ANSI B31.1 code for pressure
- piping, 1983 edition which specifies a socket weld leg size of greater than 1,09 times the nominal pipe wall thickness, This acceptance criteria was applied in anticipation that other ongoing AEPSC activities directed toward a reconciliation of our original design codes versus later code editions and code cases would lead to adoption of the B31.1-1983 criteria for socket weld sizes.
The socket weld inspection was performed on various, accessible small bore piping systems in both units.
These systems were considered to be typical in general of small bore piping installed in the plant in that the socket welds in the inspected systems would have been installed using the same procedures applied to socket welds in other small bore piping throughout the plant.
In Unit 1, a total of 60 socket welds were inspected.
All inspected welds in Unit 1 were within the established acceptance criteria.
In Unit 2,. an initial sample of 60 welds were inspected, with 6 welds initially identified as falling below the acceptance criteria.
Consequently, an,expan'ded sample of 200 welds in Unit 2 were inspected with an additional 4 welds falling below the acceptance criteria.
In all 10 cases where initial data indicated a weld size less than the acceptance
- criteria, a system operability determination was made by plant personnel and confirmed by AEPSC design within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
All systems in which potentially undersized welds were identified were det'ermined to be operable and no compromise of plant safety occurred.
Subsequent investigations have found that 2 of the 6
initially identified non-conforming welds in Unit 2 were actually within the established acceptance
- criteria, Preliminary evaluation of the remaining 8 welds indicates that the degree to which they are undersized versus the B31.1-1983 requirement is small enough that the systems in which they are installed will continue to perform their design function and will not impact plant safety.
The
Mr. A. B. Davis AEP'NRC:1090H limiting dimensions (leg and throat sizes) on all these welds were within 1/32 inch of the acceptance criteria.
All observed discrepancies have been included in our problem report system.
Our evaluation of the root cause of the observed discrepancies and identification of any required corrective actions are continuing.
No further inspections of socket welds are planned at this time.
We are,
- however, continuing with more.detailed evaluations of plant as-found data to confirm the conclusions given above.
As previously agreed with Region III, we will continue to inform you of any new information arising from these evaluations, We are reviewing appropriate
- codes, code cases, etc. for applicability to the results of the socket weld inspections and are continuing the code reconciliation activities mentioned above.
Any licensing or regulatory issues arising from these efforts will be addressed as required.
2.
Desi n Control Corrective Actions Our submittal of June 2, 1989, (AEP:NRC:1060N) provided a
description of the actions that have been initiated in conjunction with other existing internal review programs to confirm the overall adequacy of our installed systems and design activities, as well as to expedite the implementation of any required improvements to our design control process.
As discussed in the June 2 submittal, a management steering committee comprised of representatives from the AEPSC Design Division, Nuclear Engineering Department, Quality Assurance Division and Nuclear Safety and Licensing Section was established to direct the implementation of these activities.
The programs initiated under the cognizance of the steering committee are intended to provide the insights necessary to establish a root cause for any identified deficiencies in our design and design control process in order to allow initiation of a definitive course of corrective action to eliminate the deficiency.
Several of the actions discussed in the above referenced June 2, 1989 response to the design control violation cited in Inspection Reports 50-315/88028 and 50-316/88032 have been initiated or completed in approximately the same time frame in which the inspection activities addressed by this response were conducted.
Specifically, a review by an independent contractor of a sample of AEPSC design calculations was completed; a detailed AEPSC Quality Assurance Division audit
Mr. A. B. Davis AEP:NRC:1090H of installed, design complete and in-design requests for change (RFCs) and minor modifications (MMs) has been completed with the assistance of an independent contractor; walkdowns of safety related large bore (greater than 2 1/2 inches) piping systems and pipe supports have begun for the purpose of development of as-built drawings and independent analysis of as-built large bore piping systems for comparison to design analyses previously performed under AEPSC cognizance.
Our assessment of the programs initiated to date is that they are well focused and should effectively lead to a determination of the root cause of any -identified deficiencies.
The evaluations of the results of completed activities, including final root cause determinations, are in progress.
Until these evaluations are complete, a definitive statement as to the root cause of any identified design process weakness cannot be made.
In the absence of a complete determination of the underlying reasons for weaknesses disclosed through our design confirmation programs, initiation of action's beyond those currently planned would be premature.
Upon completion of the evaluations of the results of our ongoing design review activities an assessment will be made of the need for an expansion in the scope of our design confirmation program.
In the interim, consistent with the commitment we have made to the Region III staff, we will continue to inform you of the progress of our design confirmation activities.
In the. specific areas of piping and pipe support design, additional actions as described in the attached response to the NRC Notice of Violation have been initiated to improve design calculation documentation.
It should also be noted that effective September 1,
1989 a reorganization in the Design Division occurred creating the Nuclear Design Group to provide focused management attention in the area of Cook'uclear Plant and nuclear design activities.
This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.
Sincerely, I
M. P. Al xich Vice President
Mr. A. B. Davis AEP:NRC:1090H eh Attachment cc:
D. H. Williams, Jr.
A. A. Blind - Bridgman R.
C. Callen G. Charnoff NFEM Section Chief A. B. Davis
- Region III NRC Resident Inspector
- Bridgman
ATTACHMENT 1 TO AEP:NRC:1090H
RESPONSE
TO VIOLATIONS
Attachment to AEP:NRC:1090H Page 1
NRC Violation 1 "10 CFR 50, Appendix B, Criterion V, as implemented by the D.
C, Cook Operations Quality Assurance Program requires that activities affecting quality shall be accomplished in accordance with prescribed procedures.
Contrary to the above, as required by D.
C.
Cook Procedure PMI-7030, Paragraph 4.5.1, a Condition Report was not initiated when design deficiencies were discovered for pipe supports 1-GFV-L812, 1-ACA-R911, 1-ACA-R910, 1-ACCW-R924, and 1-GCCV-L181 (315/89002-01; 316/89022-01).
This is a Severity Level IV violation (Supplement 1)."
Response
to Violation At the time the design discrepancies were identified, the primary effort was focused on completing the required piping system stress
- analysis, taking into consideration the additional identified support discrepancies, to complete the investigation and identification of required corrective action to allow re-start of the unit.
Documentation of the discrepancies was not placed in the problem report/problem report trending system, (1)
Corrective Actions Taken and Results Achieved Problem reports for the discrepant supports have been initiated for inclusion in the problem report/problem report trending system for Cook Nuclear Plant.
(2)
Corrective Action Take to Avoid Furthei Violation Pipe support discrepancies found during the establishment of the as-installed configuration of the piping system are factored into the pipe stress analysis.
In order to ensure that these discrepancies are documented through our problem report system, the analysis verification checklist will be revised to include an explicit review prior to close-out of the analysis package to determine if the initiation of additional problem reports is required.
Attachment to AEP:NRC:1090H Page 2
(3)
Date When Full Com liance Will be Achieved Problem reports for the discrepant pipe supports in the cited violation were" initiated on August 25, 1989 (1-GCCW-L181), August 31, 1989 (1-GFW-L812) and October 4,
1989 (1-ACA-910 & 911 and 1-ACCW-R924).
The revision to the analysis verifi'cation checklist will be completed by October 31, 1989.
NRC Violation 2 "10 CFR 50, Appendix B, Criterion III, as implemented by the D.
C.
Cook Operations Quality Assurance Program requires that measures be established to check and verify the design.
Contrary to the above, the design was not adequately checked and verified in that:
a.
For pipe support 1-CCW-L181, the bending stress in the redesigned support was not evaluated (315/89022-03A; 316/89022-03A).
b.
For pipe support 1-CS-R541, two of four load vectors on the support were not evaluated (315/89022-03B; 316/89022-03B).
This is a Severity Level IV violation (Supplement 1)."
Response
to Violation:
a)
During the calculation process the bending stress at the shear pin was judged to be acceptable, however, this engineering judgement was not documented in the calculation package.
A confirmatory calculation was performed during the inspection showing that the bending stresses are within the design criteria limits.
b)
During the calculation process, all four load conditions were considered with the thermal load determined to be the condition dictating which vectors were to be used for the calculation.
The critical load vectors were subsequently used in the design,
- however, the calculation package did not document the engineering judgement-that the other load vector combinations were non-critical.
Attachment to AEP:NRC:1090H Page 3
(1)
Corrective Actions Taken and Results Achieved Preliminary confirmatory calculations were performed at the time of the inspection and indicated that the supports are within the design criteria limits.
(2)
Corrective Action Taken to Avoid Further Violation Standard calculation forms for pipe supports are being developed along with a revision to the checklists to be used during the checking process.
The revised checklist and standardized forms will explicitly assure that assumptions and engineering judgements associated with pipe support calculations have been documented and verified.
(3)
Date When Full Com liance Will be Achieved The calculations will be updated to reflect the above
- concerns, standard calculation forms will be developed, and a checklist revision completed by October 31, 1989.