ML18022A542

Revises 870702 Response to Violations Noted in Insp Rept 50-400/87-77.Util in Progress of Closing Out Remaining Structures,Including Rcb Platforms,Rab 248 Platform & Steam Generator Lower Lateral Supports.Final Response by 871101

ML18022A542
Person / Time
Site:	Harris
Issue date:	08/04/1987
From:	Watson R CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
CON-NRC-569 HO-870473-(O), NUDOCS 8708070205
Download: ML18022A542 (9)

Text

I It l REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

¹ II ACCESSION NBR 8708070205 DQC. DATE: 87/08/04 NOTARIZED: NO DOCKET FAC IL: 50-400 Shearon Harris Nuclear Power Planti Unit 1> Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION WATSON> R. A. Carolina Pacer 5 Light Co.

RECIP. NAME RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Revises 870702 response to violations noted in Insp Rept 50-400/8f-77. Car rec tive actions: closing out of calculations for remaining stY uctures in progress'ncluding RCB platforms 5 RAB 248 platform.

DISTRIBUTION CODE: IEOID COPIES RECEIVED: LTR I ENCL g SIZE:

TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES: Application f or p ermi t r eneuja 1 f i l ed. 05000400 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 BUCKLEY'S B 2 2 INTERNAL: ACR S 1 AEQD 1 1 DEDRO 1 1 NRR MORISSEAUi D 1 1 NRR/DOEA DIR 1 NRR/DREP/EPB 1 1 NRR/DREP/RPB 2 2 NRR/DRIS DIR 1 1 NRR/PMAS/ ILRB 1 1 0 ERMAN> J 1 OGC/HDS1 1 1 1 RES DEPY GI 1 1 RGN2 FILE 01 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 22

CITAL Carolina Power & Light Company HARRIS NUCLEAR PROJECT P. 0 ~ Box 165 New Hill, NC 27562 P,UG O g >987 File Number'. SHF/10-13510E Letter Number'HO-870473 (0) NRC-569 Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555 Gentlemen'.

Carolina Power 6 Light Company wishes to revise the reply dated July 2, 1987 concerning Violation "A" identified in IE Report 50-400/86-77. Further review and discussion of the violation circumstances indicates the need to correct statements included in our initial response. Changes to the response are identified by a line in the right hand margin of the affected page.

A final response to this item is still projected to be issued by November 1, 1987 as stated in our initial response.

Thank you for your consideration in this matter.

Very truly yours, R. A. Watson Vice President Harris Nuclear Project RAW:lkd Attachment t cc'Messrs.

Dr. J.

8708070205 870804 PDR ADOCK 05000400 PDR B. C. Buckley (NRC)

G. Maxwell (NRC-SHNPP)

Nelson Grace (NRC)

MEM/HO-8704730/PAGE 1/OS1 Q~)0

0 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 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-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

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pool rack support 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'.

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-8704730/PAGE 2/Osl

0 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 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 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. 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. 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 CPSL 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-8704730/PAGE 3/OS1

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-8704730/PAGE 4/OS1