ML17250A668
| ML17250A668 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/25/1980 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| IEB-79-14, NUDOCS 8010020113 | |
| Download: ML17250A668 (4) | |
Text
.e pip Piit/81 (,tI'iIiiiI'I'l3 ROCHESTER GAS AND ELECTRIC CORPORATION
'e ral ~ 4' Irrr 50 4 IC 55455
~
89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE. JR.
V5CC PRCSIDCNT August 25, 1980 TC5 CPHONC AREA CODS 75S 54I6.2700 Mr. Boyce H. Grier, Director U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue IQng of Prussia, Pennsylvania 19406
Subject:
1E Bulletin No. 79-14 Seismic Analyses for As-Built Safety Related Piping Systems R. E. Ginna Nuclear Power Plant, Unit No.
1 Docket No. 50-244
Dear Mr. Grier:
On October 30, 1979 we provided you with the 120 day report required in the subject bulletin.
The purpose of that letter was to provide the results of the inspections to verify that seismic analysis input informa-tion conforms to the actual configuration of safety-related systems.
There were certain sections for which the necessary information was not available, as stated-in that letter.
The purpose of this letter is to provide the infor-mation which remained, obtained during the spring refueling maintenance outage.
The first group of supports involved approximately 40 pipe supports where we were unable to determine some weld and lug sizes or spring rates prior to the refueling maintenance outage.
After additional attempts all pre-viously unverified information was obtained except in the case of those embedded in concrete.
Alldata was evaluated and the supports determined to be acceptable.
The list of discrepancies and each evaluation is provided in Supplemental Report LER 79-015/01X-1 under a separate letter.
The second group involved seal water return piping and charging pump suction piping located in the Spent Resin Storage Tank Rooms.
An ALARAevaluation was performed to assess the potential radiation exposure for any rework of the seven supports involved, and determined the exposures to be unacceptably high.
Because of this an investigation was made to determine the effects of the failure of these supports..An analysis was performed with the following assumptions:
801010 I/Q
p.
ee
ROCHESTER GAS ANP ELECT CORP.
pATE August 25, 1980 To Mr. Boyce H. Grier, Director SHEET NO.
l.
One support, SWCH-16, found to be acceptable, is active.
E
~ 2.
Supports SWCH-17, 67, 68, 38, 76 and 77 are inactive.
3.'n anchor exists at the north wall of the Spent Resin Tank Room, since the pipe sleeves are filled with concrete.
1 4.
The design thermal temperature is 500 F.
5.
The system was analyzed in accordance with the IE Bulletin 79-14 evaluation and acceptance criteria.
It was concluded from the analysis that the only area of potential concern is due to the calculated stress resulting from the combined stresses with the safe shutdown earthquake (SSE) stress.
This exceeds the code allowable for an SSE event, but is less than the pipe yield stress, and loss of integrity willnot occur.
Therefore, failure of supports SWCH-17, 67, 68, 38, 76 and 77 willnot create an unsafe condition, and it is not necessary to subject personnel to the high radiation dosage that would be incurred during any rework of those supports.
Very truly yours, L. D. White, Jr.
xc; U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C.
20555