ML18005A192
| ML18005A192 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/30/1987 |
| From: | Watson R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| HO-870533O, NUDOCS 8711030511 | |
| Download: ML18005A192 (6) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (R IDS)
ACCESSION NBR: 8711030511 DQC. DATE: 87/10/30 NOTARIZED:
NO FACIL: 50-400 Shear on Harris Nuclear Power Planti Unit ii Carolina AUTH. NAME AUTHOR AFFILIATION MATSONi R. *.
Carolina Poeer 5 Light Co.
REC IP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
DOCKET ¹ 05000400
SUBJECT:
Revised response to violations noted in Insp Rept 50-400/86-77,per 871030 telcon. Corrective actions: spent fuel pool rack support design changed to eliminate questionable supports h substitute surface mounted bearing plates.
DISTRIBUTION CODE:
IE01D COPIES RECEIVED: LTR ENCL SIZE:
TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES: Application for permit renewal filed.
05000400 REC IP IENT ID CODE/NAME PD2-1 PD INTERNAL:
AEOD NRR MORISSEAU, D NRR/DREP /EP 8 NRR/DRIS DIR O
MANiJ REG FILE RG FILE 01 EXTERNAL:
LPDR NSIC CQP IES LTTR ENCL 1
1 RECIPIENT ID CODE/NAME BUCKLEYr B DEDRQ NRR/DOEA DIR NRR/DREP/RPB NRR/PMAS/ ILRB OGC/HDS1 RES DEPY GI NRC PDR COP IES LTTR ENCL 2
2 1
1 1
2 2
1 1
1 1
1 1
1 1
TOTAL NUMBER QF CQP IES REQUIRED:
LTTR 20 ENCL 20
C~L Carolina Power & Light Company HARRIS NUCLEAR PROJECT P. 0.
Box 165 New Hill, North Carolina 27562 OCT 50 l987 File Number'.
SHF/10-13510E Letter Number.'0-8705330 Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555 NRC-587 Gentlemen.'ursuant to our telephone call with your Mr.
D.
M. Verrelli on October 30,
- 1987, regarding our reply to Violation "A" identified in IE Report 50-400/86-77, the attached is Carolina Power 6 Light Company's revised response.
It is considered that the corrective actions taken/planned are satisfactory for resolution of the item.
Thank you for your consideration in this matter.
Yours very truly, R. A. Watson Vice President Harris Nuclear Project RAW:acm Enclosure cc.'Mr. G. F. Maxwell (NRC-SHNPP)
Mr. B.
C. Buckley (NRC)
Dr. J.
N. Grace (NRC) 87ii03051i 87i030 PDR ADOCK 05000400 G
PDR MEM/HO-8705330/1/Osl soo~
il'
Attachment to CP&L Letter of
Response
to NRC I.E.
Report RII:
50-400/86-77 Re orted Violation:
A.
10 CFR 50, Appendix B, Criterion III, as implemented by the CP&L accepted QA 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-SOl 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 support system were not identified during the design verification process')
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:
The violation is correct as stated 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 assumed 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 length of the embed welds.
In this calculation (FCR-AS-10360) a supervisor, among
- others, was involved in the design verification process in MEM/HO-8705330/2/OS1
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.
The design verification did fail to correct design assumptions which were subsequently found to be outside code allowables, and in one instance at least, violated HPES site procedures.
The program (i.e., established procedures) is still felt to be acceptable.
Corrective Ste s Taken and Results Achieved:
Investigation into the adequacy of Detail
'G'onnections indicated 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.
Closing out of calculations for remaining structures, including the RCB platforms, RAB 248 platform, and the steam generator Lower lateral supports is complete.
However, the review to confirm that other non-code specific redesign methods were not used is continuing.
We estimate completion of this review by March 31, 1988.
Corrective Ste s Taken to Avoid Further Noncom liance:
Actions have been taken to preclude similar non-conformances in the future.
HPES Civil/Structural design guidelines were reviewed for any needed correction as a result of Detail "G" issues.
No revisions were deemed necessary.
HPES Supervisors and Design Personnel have been instructed on the procedural requirements for independence in Design verification, and AISC code interpretation.
The HPES Manual of Instructions has been revised to provide specific design verification guidelines in Procedure 3.29.
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.
One of the two supervisors involved with the improper design verification and approval was a
contract employee and has since been released.
The other supervisor, a
CP&L employee (currently assigned to the Corporate Nuclear Engineering Department) has been counseled on the procedural requirements for design verification and approval.
If the results of our remaining investigation dictate additional measures, these will be addressed in our final report.
MEM/HO-8705330/3/Osl
Date When Full Com liance Will Be Achieved:
Full compliance is pending additionaL reviews as stated above.
It is estimated that this review will be completed and a
final response submitted by March 31, 1988.
MEM/HO-8705330/4/OS1