ML18004B871
| ML18004B871 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 07/02/1987 |
| From: | Watson R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| CON-NRC-564 HO-870454-(O), NUDOCS 8707080348 | |
| Download: ML18004B871 (6) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
'CCESSION NBR: 8707080348 DOC. DATE: 87/07/02 NOTARIZED:
NO F*CIL:50-400 Shearon Harris Nuclear Poeer Planti Unit iI Carolina AUTH. NAME AUTHOR AFFILIATION WATSON'. A.
Carolina Pomer 5 Light Co.
RECIP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET (l
05000400
SUBJECT:
Responds to NRC 870602 ltr re violations noted in Insp Rept 50-400/86-77. Corrective actions: 30 of 54 Detail G
connections reinforced.
DISTRIBUTION CODE:
IEQID COPIEE RECEIVED: LTR
~
ENCL +
SIZE:~
TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES: Application for permit renewal filed.
05000400 RECIPIENT ID CODE/NAME PD2-1 PD RES
'TERNAL:
LPDR NSIC INTERNAL:
ACRS DEDRO NRR ROEI M. L NRR/DREP/EPB NRR/DR IS DIR OGC/HDS1 DEPY GI COPIES LTTR ENCL 1
2 2
1 1
1 1
1 1
1 1
1 1
1 RECIPIENT ID CODE/NAME BUCKLEYIB AEOD ENF LIEBERMAN NRR/DOEA DIR NRR/DREP/RPB N
/ILRB REG FILE ILE 01 NRC PDR COPIES LTTR ENCL 1
1 1
1 1
2 2
1 1
1 1
1 1
TOTAL NUMBER OF COPIES REQUIRED:
LTTR 22 ENCL 22
ggg Carolina Power & Light Company HARRIS NUCLEAR PROJECT P. 0.
Box 165 New Hill, NC 27562
)UL 0 2
$987 File Number'SHF/10-13510E Letter Number'HO-870454 (0)
NRC-564 Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555 Gentlemen:
In reference to your letter of June 2,
1987, referring to I.E.
Report RII:
50-400/86-77, the attached is Carolina Power
& Light Company's interim reply to violation "A" identified in Enclosure 1.
It is considered that the corrective actions taken/planned are satisfactory for resolution of the item. It is projected that a
final response to this item will be submitted by November 1,
1987.
Thank you for your consideration in this matter.
Very truly yours, R. A. Watson Vice President Harris Nuclear Project RAW:skm Attachment cc:
Messrs.
B. C. Buckley (NRC)
G. Maxwell (NRC-SHNPP)
Dr. J. Nelson Grace (NRC) p 0>48 S7070P 8
DOCK 05000400 PDR MEM/HO-8704540/PAGE 1/OS1
Attachment to CP&L Letter of Response to NRC I.E. Report RII:
50-400/86-77 Re orted Violation.'.
10 CFR 50, Appendix B, Criterion III, as implemented by th'e CP&L accepted gA program (FSAR Chapter 17.2), requires that design control measures provide for verifying or checking the adequacy of design, such as by the performance of design
- reviews, by use of alternative or simplified calculation
- methods, or by performance of a suitable testing program.
Contrary to the above, the licensee's design verification program was not adequately implemented in that:
1)
The inadequate design methodology used in the original design calculations for design of generic Detail G
connection on Drawing CAR 2168-G-251-S01 and for the calculations for Field Modification FM-C-CAR 2168-G-251-S01 were not identified during the design verification process.
2)
Incorrect application of the AISC Ultimate Strength Method for weld design and use of incorrect allowable weld stress values in calculations for Field Change Request (FCR) AS-10360 for modification of the new fuel pool rack su'pport system were not identified during the design verification process.
3)
Use of an individual who had specified the design approach and had supervisory responsibility for the individuals performing the design to verify portions of the calculations for FCR AS-10360 in violation of CP&L Procedure 3.3, Design Verification.
This is a Severity Level IV violation (Supplement II).
Denial or Admission and Reason for The Violation.'he violation is correct as stated.
I The violation occurred because of an error in the design assumptions made for distribution of weld stresses under specific connection types. and loading conditions.
In the case of Detail G
on Containment Building Cable Tray Supports, the concentrated loading at the heel was ass'umed by the engineer to distribute along the horizontal portion of the weld.
This resulted in a local, isolated overstress of the weld segment assumed to be a distance of k+t from the angle heel.
In the case of the fuel pool floor, the beam-to-embed welds were subjected to thermal stresses which were assumed to redistribute along the 1'ength of the embed welds.
In this calculation (FCR-AS-10360) a supervisor, among
- others, was involved in the design verification process in MEM/HO-8704540/PAGE 2/Osl
violation of HPES design verification procedures.
This was due to the fact that the calculations package was large and completed over a long period of time. It should be noted that while the design verification failed to correct design assumptions which were subsequently found to be outside code allowables, and in one instance at least, violated HPES site procedures, the ANSI design verification guidelines were not violated.
In particular, the ANSI standard allows the supervisor to perform design verification if the analysis is for a special area where other qualified personnel are not available.
The HPES design verification program is not, therefore, felt to be deficient.
Corrective Ste s Taken and Results Achieved:
Investigation into the adequacy of Detail 'G'onnections
'ndicated they were capable of carrying design loads without failing', however, to assure no outstanding safety issues remained and to provide margin for future plant modifications, 30 of the 54 connections were reinforced.
Details were provided on Field Modifications (FM) FM-C-11020,
- 11022, 11023,
- 11025, 11028,
- 11029, 11030,
- 11033, 11039, '11040,
- 11043, 11048-54, and 11056-67, and this work has been completed.
Similarly, spent fuel pool rack support designs have been changed per Plant Change Request (PCR) 1857 to eliminate the questionable supports and substitute surface mounted bearing plates.
Design for new supports in the new fuel. pools is in progress.
In samples and audits to date, no other examples of design verifications in violation of design control procedures have been found.
We have no evidence to indicate this is not an isolated incident.
Closing out of calculations for remaining structures is in
- progress, including the RCB platforms, RAB 248 platform, and the steam generator lower lateral supports.
These calculations are being reviewed to confirm that Detail 'G'ype connections and/or other non-code specific design methods were not used elsewhere.
We estimate completion of this review by November 1,
1987.
Corrective Ste s Taken to Avoid Further Noncom liance:
Actions have been taken to preclude similar non-conformances in the future.
In particular, HPES design guidelines have been revised to eliminate Detail 'G'nd HPESSupervisors and Design Personnel have been instructed on the procedural requirements for independence in Design verification, and AISC code interpretation.
In addition, with the completion of the construction and testing of the plant, the number of people involved with structural design has sharply decreased along with the scope of work, thereby increasing the level of management and supervisory oversight.
The supervisor involved with the improper MEM/HO-8704540/PAGE 3/Osl
design verification and approval was a contract employee and has been released.
If the results of our remaining investigation dictate additional measures, these will be addressed in our final report.
Date When Full Com liance Will Be Achieved:
Full compliance is pending closeout and review of additional calculations as stated above.
It is estimated that this review will be completed and a final response submitted by November 1,
1987.
MEM/HO-8704540/PAGE 4/OS1
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